ULNRC-05744, Transmittal of Application for Amendment of License or Construction Permit - Revision of Technical Specification 3.3.8: Difference between revisions

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{{Adams
#REDIRECT [[ULNRC-05744, Attachment 1 to 1/19/2012 Email]]
| number = ML103470204
| issue date = 12/10/2010
| title = Transmittal of Application for Amendment of License or Construction Permit - Revision of Technical Specification 3.3.8
| author name = Maglio S
| author affiliation = Ameren Missouri
| addressee name =
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000483
| license number = NPF-030
| contact person =
| case reference number = LDCN 10-0036, ULNRC-05744
| document type = Letter type:ULNRC, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 38
}}
 
=Text=
{{#Wiki_filter:WAmeren MISSOURI December 10,2010 ULNRC-05744 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop PI-137 Washington, DC 20555-0001 Ladies and Gentlemen:
DOCKET NUMBER 50-483 CALLAWAY PLANT UNION ELECTRIC CO. 10 CFR 50.90 APPLICA TION FOR AMENDMENT TO FACILITY OPERATING LICENSE NPF-30 (LDCN 10-0036) REVISION OF TECHNICAL SPECIFICATION 3.3.8 Callaway Plant Pursuant to 10 CFR 50.90, "Application for Amendment of License or Construction Permit," Ameren Missouri (Union Electric Company) herewith transmits an application for amendment to Facility Operating License Number NPF-30 for the Callaway Plant. The proposed amendment would add new Surveillance Requirement (SR) 3.3.8.6 to Technical Specification (TS) 3.3.8, "Emergency Exhaust System (EES) Actuation Instrumentation." The new SR would require the performance of response time testing on the portion of the EES required to isolate the normal fuel building ventilation exhaust flow path and initiate the fuel building ventilation isolation signal (FBVIS) mode of operation.
Attachments 1 through 5 provide the Evaluation, Markup of Technical Specifications, Retyped Technical Specifications, Proposed Technical Specification Bases Changes, and Proposed FSAR Changes, respectively, in support of this amendment request. Attachments 4 and 5 are provided for information only. Final TS Bases Changes will be processed under Callaway's program for updates per TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is implemented.
The FSAR will be updated under the normal update process pursuant to 10 CFR 50.71(e).
No commitments are contained in this amendment application.
PO Box 620 Fulton, MO 65251 AmerenMissouri.com VLNRC-05744 December 10, 2010 Page 2 It has been determined that this amendment application does not involve a significant hazard consideration, as determined per 10 CFR 50.92, "Issuance of Amendment." Pursuant to 10 CFR 51.22, "Criterion for Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusion or Otherwise not Requiring Environmental Review," Section (b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
The Callaway Onsite Review Committee and a subcommittee of the Nuclear Safety Review Board have reviewed and approved the proposed changes and the attached licensing evaluations and have approved the submittal of this amendment application.
Ameren Missouri requests approval of this license amendment request prior to October 1, 2011 so that it can be implemented prior to the next refueling outage (Refuel 18, October 2011). Ameren Missouri further requests that the license amendment be made effective upon NRC issuance, to be implemented within 90 days from the date of issuance with the following exception:
Since SR 3.3.8.6 is a new Surveillance Requirement, the first required performance will come due by the end of the first surveillance interval that begins or is in effect on the date of implementation of this amendment.
This is similar to the License Condition applied to new Surveillance Requirements added by License Amendment 133 for the ITS Conversion.
As such, if the license amendment is issued prior to October 1, 2011, SR 3.3.8.6 will first be met during Refuel 18. If the license amendment is issued after October 1, 2011, SR 3.3.8.6 will first be met during Refuel 19 (spring 2013). In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," Section (b )( 1), a copy of this amendment application is being provided to the designated Missouri State official.
If you have any questions on this amendment application, please contact me at (573) 676-8719 or Mr. Tom Elwood at (314) 225-1905.
I declare under penalty of perjury that the foregoing is true and correct. Executed on: l 'L It D /1-0 l b GGY/nls Very truly yours, J.) t<>ti A. Scott A. Maglio ./ Regulatory Aff'a(rs Manager ULNRC-05744 December 10, 2010 Page 3 Attachments 1 -Evaluation 2 -Markup of Technical Specifications 3 -Retyped Technical Specifications 4 -Proposed Technical Specification Bases Changes (for information only) 5 -Proposed FSAR Changes (for information only)
ULNRC-05744 December 10, 2010 Page 4 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy) Attn: Document Control Desk Washington, DC 20555-0001 Mr. Elmo E. Collins, Jr. Regional Administrator U.S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mohan C. Thadani (2 copies) Senior Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8G 14 Washington, DC 20555-2738 Mr. James Polickoski Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8B 1 A Washington, DC 20555-2738 ULNRC-05744 December 10, 2010 Page 5 Index and send hardcopy to QA File A160.0761 Hardcopy:
Certrec Corporation 4200 South Hulen, Suite 422 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed).
Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:
A. C. Heflin F. M. Diya L. S. Sandbothe C. O. Reasoner III S. A. Maglio S. L. Gallagher T. L. Woodward (NSRB) T. B. Elwood G. G. Yates Ms. Diane M. Hooper (WCNOC) Mr. Tim Hope (Luminant Power) Mr. Ron Barnes (APS) Mr. Tom Baldwin (PG&E) Mr. Wayne Harrison (STPNOC) Ms. Linda Conklin (SCE) Mr. John O'Neill (Pillsbury, Winthrop, Shaw, Pittman LLP) Missouri Public Service Commission Mr. Dru Buntin (DNR)
Attachment 1 Page 1 of9 EVALUATION
: 1. DESCRIPTION
: 2. PROPOSED CHANGES 3. BACKGROUND
: 4. TECHNICAL ANAL YSIS 5. REGULATORY SAFETY ANALYSIS 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION 5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA
: 6. ENVIRONMENTAL CONSIDERATION
: 7. REFERENCES Page 2 Page 2 Page 2 Page 3 Page 4 Page 5 Page 7 Page 9 Page 9 Attachment 1 Page 2 of9
 
==1.0 DESCRIPTION==
 
EVALUATION The proposed amendment would add new Surveillance Requirement (SR) 3.3.8.6 to Technical Specification (TS) 3.3.8, "Emergency Exhaust System (EES) Actuation Instrumentation." The new SR would require the performance of response time testing on the portion of the EES required to isolate the normal fuel building ventilation exhaust flow path and initiate the fuel building ventilation isolation signal (FBVIS) mode of operation.
 
==2.0 PROPOSED CHANGE==
S This amendment application requests the addition of new SR 3.3.8.6 which would read as follows: "Verify Fuel Building Ventilation Exhaust ESF RESPONSE TIMES are within limits." The Surveillance Frequency will be (at least once per) 18 months on a STAGGERED TEST BASIS. New SR 3.3.8.6 will have a NOTE excluding the radiation monitor detectors from response time testing. TS Table 3.3.8-1 will also be revised to indicate that new SR 3.3.8.6 applies to automatic actuation Function 2, "Automatic Actuation Logic and Actuation Relays (BOP ESF AS)," and Function 3, "Fuel Building Exhaust Radiation
-Gaseous." The Surveillance Frequency, SR Note, and the TS Table 3.3.8-1 changes are identical to those approved by the NRC in Reference 7.1 for Control Room Ventilation Isolation.
The TS markups and retyped pages are provided in Attachments 2 and 3, respectively.
Corresponding TS Bases changes and FSAR changes are provided for information only in Attachments 4 and 5.
 
==3.0 BACKGROUND==
 
The fuel building at Callaway is served by an outside air supply system which provides fresh outside air, either heated or cooled as required, to all areas of the fuel building.
The supply air unit has provisions for operating in a recirculation mode. Within the fuel building, the auxiliary/fuel building normal exhaust system takes suction from the area above the spent fuel pool and mixes that air with the air from the auxiliary building prior to processing it through the auxiliary/fuel building filter adsorber train and discharging it to the unit vent.
Attachment 1 Page 3 of9 The emergency exhaust system (EES) collects and processes the fuel building atmosphere in the event of a fuel handling accident.
During operation of the EES, the fuel building non-essential heating, ventilation, and cooling (HV AC) air paths are isolated and the building exhaust is processed through safety grade filter-adsorber units to assure that fission products and particulate matter are collected and processed.
The fuel building intake air system is provided with two motor-operated dampers in a series arrangement. Each damper is powered from a separate Class 1 E source to assure closure. Transfer from the normal HV AC operations to the emergency HV AC operations occurs automatically upon receipt of a fuel building ventilation isolation signal (FBVIS). The EES maintains a minimum negative pressure of 114 in. w.g. to assure that all leakage is into the building.
The EES is on standby for an automatic start following receipt of a FBVIS or a safety injection signal (SIS). The initiation of the SIS mode of operation (such as would be the case after a LOCA) takes precedence over the FBVIS mode of operation; however, this amendment request is concerned only with the FBVIS mode. Actuation of the emergency mode of operation is initiated either manually by operator action or automatically upon detection of high radiation levels in the fuel building ventilation exhaust. Actuation of the FBVIS isolates the outside air intake system, trips the supply air handling units, and closes the corresponding dampers in the normal exhaust ductwork to the auxiliary building in order to isolate the fuel building.
The accident analysis for a fuel handling accident in the fuel building in FSAR Section 15.7.4 relies on automatic actuation of the EES and realignment to the FBVIS mode of operation to mitigate the radiological consequences of such an event. 4.0 TECHNICAL ANALYSIS A fuel handling accident (FHA) may be postulated to occur during movement of irradiated fuel assemblies within the fuel building.
In the fuel building, a fuel assembly could be dropped in the transfer canal, in the fuel storage pool, or in the cask loading pool. In addition to the area radiation monitors located on the wall around the fuel storage pool, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations.
The doors in the fuel building are closed to maintain controlled leakage characteristics in the fuel storage pool region during operations involving irradiated fuel. Should a fuel assembly be dropped in the canal, in the cask loading pit, or in the pool and release radioactivity above a prescribed level, the radiation monitors sound an audible alarm. If one of the redundant fuel building ventilation exhaust radiation monitors, GGRE0027 or GGRE0028, indicates that the radioactivity in the exhaust is greater than the set limits, an alarm sounds and the auxiliary/fuel building normal exhaust is switched to the emergency exhaust system (EES) to allow the spent fuel pool ventilation to exhaust through the engineered safety feature (ESP) charcoal filters to remove most of the halogens prior to discharging to the atmosphere via the unit vent. The normal ventilation Attachment 1 Page 4 of9 supply servicing the spent fuel pool area is automatically shut down, thus ensuring controlled leakage to the atmosphere through the charcoal adsorbers.
In the analysis of the radiological consequences of the fuel building FHA, it is assumed that fuel building ventilation is switched from the auxiliary/fuel building normal exhaust system to the EES within an assumed time interval beginning from the time when radioactivity in the exhaust reaches the exhaust duct and produces a count rate from the detector that corresponds to the actuation setpoint for the radiation monitor. The assumed time interval, i.e., the Fuel Building Ventilation Exhaust ESF response time, includes delays associated with generating the FBVIS (radiation monitor RM-80 micro-processor and associated electronics, inputs to and outputs from the Balance of Plant (BOP) Engineered Safety Feature (ESF) Actuation System (BOP ESFAS)) and delays associated with the change of state for the actuated components (i.e., EES exhaust fan spin-up and damper stroke times). The activity released before completion of the fuel building ventilation switchover is assumed to be discharged directly to the environment with no credit for filtration or dilution.
The calculated dose consequences for the fuel handling accident in the fuel building are discussed in FSAR Section 15.7.4 and reported in FSAR Table 15.7-8. Based on historical EES damper stroke times measured under Callaway's safety-related preventive maintenance (PM) program, as well as response time testing performed per SR 3.3.7.6 for similar design elements associated with the control room emergency ventilation system, a 90-second response time is proposed as an appropriate limit for the Fuel Building Ventilation Exhaust ESF response time, which is to be added to FSAR Table 16.3-2 per Attachment
: 5. Because the fuel building fuel handling accident analysis had not previously been performed with an assumed Fuel Building Ventilation Exhaust ESF response time of 90 seconds, reanalysis of this event resulted in small increases in the calculated dose consequences.
The new/recalculated values are reflected on the mark-up for Table 15.7-8, which is provided in Attachment
: 5. The radiological consequence increases reported in Table 15.7-8 are much less than the upper limit for a minimal increase under 10 CFR 50.59(c)(2)(iii) and NEI 96-07, Revision 1, November 2000. [An increase in dose consequence is not more than minimal if the increase (1) is less than or equal to 10% of the difference between the current calculated dose value and the regulatory guideline value (10 CFR 100 or GDC 19, as applicable), and (2) the increased dose does not exceed the current SRP guideline value for the particular design basis event.] The regulatory limits for offsite dose consequences resulting from a design basis accident are 300 rem thyroid and 25 rem whole body (10 CFR 100.11), and the Standard Review Plan (SRP) 15.7.4 limits are 75 rem thyroid and 6.25 rem whole body. As can be seen from examination of the values shown on the mark-up of Table 15.7-8, all dose increases from reanalysis of the fuel handling accident using the 90-second Fuel Building Ventilation Exhaust ESF response time are fractional, i.e., well less than one percent.
Attachment 1 Page 5 of9 Additional discussion of the functioning of the nonnal auxiliary/fuel building exhaust system and the ESF function of the EES may be found in FSAR Sections 7.3.3, 9.4.2, 9.4.3, 11.5.2.3.2.3, and 12.3.4.2.2.2.8.
5.0 REGULATORY SAFETY ANALYSIS This section addresses the standards of 10 CFR 50.92 as well as the applicable regulatory requirements and acceptance criteria.
The proposed amendment would add new Surveillance Requirement (SR) 3.3.8.6 to Technical Specification (TS) 3.3.8, "Emergency Exhaust System (EES) Actuation Instrumentation." The new SR would require the performance of response time testing on the portion of the EES required to isolate the nonnal fuel building ventilation exhaust flow path and place the EES in the fuel building ventilation isolation signal (FBVIS) mode of operation.
5.1 No Significant Hazards Consideration (NSHC) Ameren Missouri has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," Part 50.92( c), as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No There are no design changes associated with the proposed change. All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable.
The proposed change will not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained with respect to such initiators or precursors.
There will be no change to fuel handling methods and procedures.
Therefore, there will be no changes that would serve to increase the likelihood of occurrence of a fuel handling accident.
The proposed change changes a perfonnance requirement, but it does not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions.
The proposed TS change will serve to assure that the fuel building ventilation exhaust ESF response time is tested and confirmed to be in accordance with the system design and consistent with the assumptions of the fuel building FHA analysis (as revised).
As Attachment 1 Page 6 of9 such, the proposed change will not alter or prevent the capability of structures, systems, and components (SSCs) to perfonn their intended functions for mitigating the consequences of an accident and meeting applicable acceptance limits. The proposed change will not affect the source term used in evaluating the radiological consequences of a fuel handling accident in the fuel building.
However, the Fuel Building Ventilation Exhaust ESF response time has been increased to 90 seconds in recognition of the total delay times involved in the generation of a fuel building ventilation isolation signal (FBVIS) and the times required for actuated components to change state to their required safety configurations.
Consequently, the fuel handling accident radiological consequences as reported in FSAR Table 15.7-8 have increased.
However, the increases are much less than the upper limit of "minimal" as defined pursuant to 10 CFR 50.59(c)(2)(iii) and NEI 96-07 Revision 1. Therefore, there is no significant increase in the calculated consequences of a postulated design basis fuel handling accident in the fuel building.
The applicable radiological dose criteria of 10 CFR 100.11,10 CFR 50 Appendix A General Design Criterion 19, and SRP 15.7.4 will continue to be met. New SR 3.3.8.6 is added to ensure system perfonnance consistent with the accident analyses and associated dose calculations (as revised).
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No With respect to any new or different kind of accident, there are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC perfonns its specified safety function.
The proposed change will not affect the normal method of plant operation or change any operating parameters.
No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment.
The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, Solid State Protection System, BOP ESF AS, MSFIS, or LSELS used in the plant protection systems. The proposed change does not, therefore, create the possibility of a new or different accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety? Response:
No Attachment 1 Page 70f9 There will be no effect on those plant systems necessary to assure the accomplishment of protection functions associated with reactor operation or the reactor coolant system. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (FQ), nuclear enthalpy rise hot channel factor (F loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other limit and associated margin of safety. Required shutdown margins in the COLR will not be changed. The proposed change does not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications.
The proposed change would add a new Technical Specification Surveillance Requirement for assuring the satisfactory performance of the fuel building ventilation exhaust ESF function in response to a FBVIS. The accident analysis for a fuel handling accident in the fuel building was performed to support the proposed Fuel Building Ventilation Exhaust ESF response time, and this reanalysis demonstrated that the acceptance criteria continue to be met with only a slight increase in radiological consequences (i.e., less than one percent).
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
 
== Conclusion:==
 
Based on the above evaluation, Ameren Missouri concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements
/ Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TSs) as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The U.S. Nuclear Regulatory Commission's (NRC's) requirements related to the content of the TSs are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items in the following specific categories:
(1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR 50.36(c)(3);
(4) design features; and (5) administrative controls.
This amendment application is related to the third category above (SRs) and is a more restrictive change since a new Surveillance Requirement is being added. The following regulatory requirements and guidance documents also apply to the EES and its actuation instrumentation:
Attachment 1 Page 8 of9
* GDC 2 requires that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without the loss of the capability to perform their safety functions.
* GDC 4 requires that structures, systems, and components important to safety be designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with the normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, discharging fluids that may result from equipment failures, and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
* GDC 13 requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.
* GDC 20 requires that the protection system(s) shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
* GDC 21 requires that the protection system(s) shall be designed for high functional reliability and testability.
* GDC 22 through GDC 25 and GDC 29 require various design attributes for the protection system(s), including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.
* Regulatory Guide 1.22 discusses an acceptable method of satisfying GDC-20 and GDC-21 regarding the periodic testing of protection system actuation functions.
These periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident.
Attachment 1 Page 9 of9
* 10 CFR 50.55a(h) requires that the protection systems meet IEEE 279-1971.
Section 4.2 of IEEE 279-1971 discusses the general functional requirement for protection systems to assure they satisfy the single failure criterion.
There will be no changes to the EES or its actuation instrumentation such that compliance with any of the above regulatory requirements and guidance documents would come into question.
The discussions in Sections 3.0 and 4.0 of this Evaluation demonstrate that the plant will continue to comply with all applicable regulatory requirements.
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 6.0 ENVIRONMENTAL CONSIDERATION Ameren Missouri has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c )(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
 
==7.0 REFERENCES==
 
7.1 Callaway License Amendment No. 152, NRC Letter dated September 9, 2002, "Callaway Plant, Unit 1 -Issuance of Amendment RE: Equipment Hatch and Emergency Air Lock Open During Core Alterations or Movement of Irradiated Fuel Assemblies Inside Containment (TAC NO. MB3605)."
ATTACHMENT 2 MARKUP OF TECHNICAL SPECIFICATIONS SURVEILLANCE REQUIREMENTS EES Actuation Instrumentation 3.3.8 ----------------------------------------------------------
NOTE ---------------------------------------------------------
Refer to Table 3.3.8-1 to determine which SRs apply for each EES Actuation Function.
SURVEILLANCE SR 3.3.8.1 Perform CHANNEL CHECK. SR 3.3.8.2 Perform COT. SR 3.3.8.3 ---------------------------------
NOTE ---------------------------
The continuity check may be excluded.
Perform ACTUATION LOGIC TEST. SR 3.3.8.4 --------------------------------
NOT E ---------------------------
Verification of setpoint is not required.
Perform TADOT. SR 3.3.8.5 Perform CHANNEL CALIBRATION.
CALLAWAY PLANT 3.3-72 FREQUENCY 12 hours 92 days 31 days on a STAGGERED TEST BASIS 18 months 18 months Amendment No. 197 I INSERT 1 SURVEILLANCE S R 3.3.8.6 -------------------------
NOT E --------------------------
Radiation monitor detectors are excluded from response time testing. FREQUENCY Verify Fuel Building Ventilation Exhaust ESF 18 months on a RESPONSE TIMES are within limits. STAGGERED TEST BASIS APPLICABLE MODES OR SPECIFIED FUNCTION CONDITIONS
: 1. Manual (a) Initiation
: 2. Automatic (a) Actuation Logic and Actuation Relays (BOP ESFAS) 3. Fuel Building Exhaust Radiation
-Gaseous EES Actuation Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 1 ) EES Actuation Instrumentation REQUIRED CHANNELS 2 2 trains 2 SURVEILLANCE REQUIREMENTS SR 3.3.8.4 SR 3.3.B.3 S'/l 3,. 3. 'i./, SR 3.3.8.1 SR 3.3.8.2 SR 3.3.B.5 :3.:3. 'i.' NOMINAL TRIP SETPOINT NA NA (b) (a) During movement of irradiated fuel assemblies in the fuel building. (b) Nominal Trip Setpoint concentration value ().1Ci/cm 3) shall be established such that the actual submersion dose rate would not exceed 4 mRlhr in the fuel building.
CALLAWAY PLANT 3.3-73 Amendment No. 197 I ATTACHMENT 3 RETYPED TECHNICAL SPECIFICATIONS EES Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS
-----------------------------------------------------------NOTE -----------------------------------------------------------
Refer to Table 3.3.8-1 to determine which SRs apply for each EES Actuation Function.
SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL CHECK. 12 hours SR 3.3.8.2 Perform COT. 92 days SR 3.3.8.3 ---------------------------------
NOT E ----------------------------
The continuity check may be excluded.
-----------------------------------------------------------------------
Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS SR 3.3.8.4 ---------------------------------
NOT E ----------------------------
Verification of setpoint is not required.
-----------------------------------------------------------------------
Perform TADOT. 18 months SR 3.3.8.5 Perform CHANNEL CALIBRATION.
18 months SR 3.3.8.6 ---------------------------------
NOTE ----------------------------
Radiation monitor detectors are excluded from response time testing. ----------------------------------------------------------------------_.
Verify Fuel Building Ventilation Exhaust ESF 18 months on a RESPONSE TIMES are within limits. STAGGERED TEST BASIS CALLAWAY PLANT 3.3-72 Amendment No. ###
APPLICABLE MODES OR SPECIFIED FUNCTION CONDITIONS
: 1. Manual (a) Initiation
: 2. Automatic (a) Actuation Logic and Actuation Relays (BOP ESFAS) 3. Fuel (a) Building Exhaust Radiation
-Gaseous Table 3.3.8-1 (page 1 of 1) EES Actuation Instrumentation EES Actuation Instrumentation 3.3.8 REQUIRED SURVEILLANCE NOMINAL TRIP CHANNELS REQUIREMENTS SETPOINT 2 SR 3.3.8.4 NA 2 trains SR 3.3.8.3 NA SR 3.3.8.6 2 SR 3.3.8.1 (b) SR 3.3.8.2 SR 3.3.8.5 SR 3.3.8.6 (a) During movement of irradiated fuel assemblies in the fuel building. (b) Nominal Trip Setpoint concentration value (f..lCi/cm
: 3) shall be established such that the actual submersion dose rate would not exceed 4 mRlhr in the fuel building.
CALLAWAY PLANT 3.3-73 Amendment No. ###
ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (for information only)
BASES ESFAS Instrumentation B 3.3.2 SURVEILLANCE SR 3.3.2.10 (continued)
REQUIREMENTS CALLAWAY PLANT Response time verification acceptance criteria are included in Reference
: 9. No credit was taken in the safety analyses for those channels with response times listed as N.A. No response time testing requirements apply where N.A. is listed in Reference
: 9. Individual component response times are not modeled in the analyses.
The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).
The safety analyses include the sum of the following response time components:
: a. Sensing circuitry delay time from the time the trip setpoint is reached at the sensor until an ESFAS actuation signal is generated by the SSPS (response time testing associated with LSELS and BOP-ESFAS is discussed under SR 3.3.5.4...aA8 SR
)------I ..... ",J. . b. Any intentional time delay set into the trip 33,U) (Ieadllag) associated with the steam line pressure high negative rate trip function) to add margin or prevent spurious trip signals; and c. The time for the final actuation devices to reach the required functional state (e.g., valve stroke time, pump or fan spin-up time). For channels that include dynamic transfer functions (e.g., lag, leadllag, ratellag, etc.), the response time verification is performed with the time constants set at their nominal values. Time constants are verified during the performance of SR 3.3.2.9. The response time may be verified by a series of overlapping tests, or other verification (e.g., Ref. 10 and Ref. 14), such that the entire response time is verified.
Response time may be verified by actual response time tests in any series of sequential, overlapping, or total channel measurements, or by the summation of allocated sensor, signal processing, and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests); (2) inplace, onsite, or offsite (e.g. vendor) test measurements; or (3) utilizing vendor engineering specifications.
WCAP-13632-P-A Revision 2, IIElimination of Pressure Sensor Response (continued)
B 3.3.2-64 Revision 8g CREVS Actuation Instrumentation B 3.3.7 B 3.3 INSTRUMENTATION B 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation BASES BACKGROUND APPLICABLE SAFETY ANALYSES CALLAWAY PLANT The CREVS provides an enclosed control room environment from which the unit can be operated following an uncontrolled release of radioactivity.
During normal operation, the Control Building Ventilation System provides control room ventilation.
Upon receipt of an actuation signal, the CREVS initiates filtered ventilation and pressurization of the control room. This system is described in the Bases for LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)." The actuation instrumentation consists of two gaseous radiation channels in the control room air intake. A high radiation signal from either of these channels will initiate both trains of the CREVS. Since the radiation monitors include an air sampling system, various components such as sample line valves and sample pumps are required to support monitor OPERABILITY.
The control room operator can also initiate CREVS trains by manual switches in the control room. The CREVS is also actuated by a Phase A Isolation signal, a Fuel Building Ventilation Isolation signal (FBVIS), or a high radiation signal from the containment purge exhaust gaseous radiation channels.
The Phase A Isolation Function is discussed in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation." The control room must be kept habitable for the operators stationed there during accident recovery and post accident operations.
The CREVS acts to terminate the supply of unfiltered outside air to the control room, initiate filtration, and pressurize the control room. These actions are necessary to ensure the control room is kept habitable for the operators stationed there during accident recovery and post accident operations by minimizing the radiation exposure of control room personnel.
In MODES 1, 2, 3, and 4, (MODE 4 is subject to LCO 3.3.2, Function 3.a), the gaseous radiation channel actuation of the CREVS is a backup for the Phase A Isolation signal actuation.
This ensures initiation of the CREVS during a loss of coolant accident or steam generator tube rupture. During CORE ALTERATIONS or during movement of irradiated fuel assemblies within containment, the gaseous radiation channel actuation of the CREVS is the primary means to ensure control room habitability in the (continued)
B 3.3.7-1 Revision Bc BASES APPLICABLE SAFETY ANALYSES (continued)
CREVS Actuation Instrumentation B 3.3.7 event of a fuel handling accident inside containment.
No control room habitablilty mitigation is required for the waste gas decay tank rupture accident.
There are no safety analyses that take credit for CREVS actuation upon high containment purge exhaust radiation.
A FBVIS is credited to protect the control room in the event of a design basis fuel handling accident bUilQing.
I,. J Verrl1/"-t1ln IS" . . . .. . , , Fuel Buildin a .. Flet"response time tested. The FBVIS for actuating a CRVIS following a Fuel Handling Accident LCO CALLAWAY PLANT in the Fuel Building. -Dt:le te lAe leeeliefJ gf tR9 F=t:lel exhaust r.di.tiQr:l R=l9Rit9FS feletioe to the Gefltf61 Ree,"" iRtalEe let:l'ver!, th9 FaVIS nill isolate tile OOlitiCl Rooill pJiOi to the aeeiaeRt QQieaeti'l8 tAe GefJtfsl Reeffi iFltalEe let:l&#xa5;8I=S a LOCA, the analysis credits a time zero Control Room isolation.
A Safety Injection signal initiates a Containment Isolation Phase A, which initiates a CRVIS. This function for isolating the Control Room prior to the post-accident radioactive plume reaching the Control Room intake louvers. For a Fuel Handling Accident within Containment, GKRE0004 and l\et:4r GKREOOOS are credited for initiating a CRVIS. These monitors C8FM8te weffi the Control Room intake louvers. They are downstream of the Control Room intake. Therefore, a specific response time is modeled, and a response time Surveillance Requirement is imposed for this CRVIS function.
The CREVS actuation instrumentation satisfies Criterion 3 of 10CFRSO.36(c)(2)(ii).
The LCO requirements ensure that instrumentation necessary to initiate the CREVS is OPERABLE.
: 1. Manual Initiation The LCO requires two channels OPERABLE.
The operator can initiate the CREVS at any time by using either of two push buttons in the control room. (continued)
B 3.3.7-2 Revision Bc EES Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation BASES BACKGROUND The EES ensures that radioactive materials in the fuel building atmosphere following a fuel handling accident are filtered and adsorbed prior to exhausting to the environment.
The system is described in the Bases for LeO 3.7.13, "Emergency Exhaust System." The system initiates filtered exhaust from the fuel building following receipt of a fuel building ventilation isolation signal (FBVIS), initiated manually or automatically upon a high radiation signal (gaseous).
High gaseous radiation, monitored by two channels, provides an FBVIS. Both EES trains are initiated by high radiation detected by either channel. Each channel contains a gaseous monitor. High radiation detected by either monitor initiates fuel building isolation, starts the EES, and initiates a CRVIS. These actions function to prevent exfiltration of contaminated air by initiating filtered exhaust, which imposes a negative pressure on the fuel building.
Since the radiation monitors include an air sampling system. various components such as sample line valves and sample pumps are required to support monitor OPERABILITY.
In the FBVIS mode, each train is capable of maintaining the fuel building at a negative pressure of less than or equal to 0.25 inches water gauge relative to the outside atmosphere.
The EES is also actuated in the LOCA (SIS) mode as described in the Bases for LCO 3.3.2, "ESFAS Instrumentation." J,.. J r /Jelerer-oce tt...r tt' S c.J,( ./'retl-n /<..1 APPLICABLE SAFETY ANALYSES LCO CALLAWAY PLANT The EES ensures that radioactive materials in the fuel building atmosphere following a fuel handling accid t are filtered and adsorbed prior to being exhausted to the environment; This action reduces the radioactive content in the fuel building exhaust following a fuel handling accident so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. ta1 control room habitability is maintained.
The EES actuation instrumentation satisfies Criterion 3 of 1 OCFR50.36( c)(2)(ii).
The LCO requirements ensure that instrumentation necessary to initiate the EES is OPERABLE. (continued)
B 3.3.8-1 Revision 8c /,
BASES (Continued)
APPLICABILITY ACTIONS CALLAWAY PLANT EES Actuation Instrumentation B 3.3.8 The manual and automatic EES initiation must be OPERABLE when moving irradiated fuel assemblies in the fuel building to ensure the EES operates to remove fission products associated with a fuel handling accident and isolate control room ventilation.
High radiation initiation of the FBVIS must be OPERABLE during movement of irradiated fuel assemblies in the fuel building to ensure automatic initiation of the EES and a CRVIS when the potential for a fuel handling accident exists. The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by unit specific calibration procedures.
Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.
This determination is generally made during the performance of a COT, when the process instrumentation is set up for adjustment to bring it within specification.
If the measured Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered. LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.
Entering LCD 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.
A second Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.
Placing a EES train(s) in the FBVIS mode of operation isolates normal air discharge from the fuel building and initiates filtered exhaust, imposing a negative pressure on the fuel building.
Further discussion of the FBVIS mode of operation may be found in the Bases for LCO 3.7.13, "Emergency Exhaust System (EES)," and in , ( continued)
B 3.3.8-3 Revision 8c BASES SURVEILLANCE REQUIREMENTS SR 3.3.B.3 (continued)
EES Actuation Instrumentation B 3.3.8 and the multichannel redundancy available, and has been shown to be acceptable through operating experience.
The SR is modified by a Note stating that the continuity check may be excluded.
This SR is applied to the balance of plant actuation logic and relays that do not have circuits installed to perform the continuity check. SR 3.3.8.4 SR 3.3.8.4 is the performance of a TADOT. This test is a check of the Manual Initiation Function and is performed every 18 months. Each Manual Initiation channel is tested through the BOP ESFAS logic. A successful test of the required contact{s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
The Frequency is based on operating experience and is consistent with the typical industry refueling cycle. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The channels tested have no setpoints associated with them. SR 3.3.8.5 A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The Frequency is based on operating experience and is consistent with the typical industry refueling cycle. _------i> " REFERENCES
::2..-4:-FSAI<. 1S:/,1-", 10CFR 100.11. CALLAWAY PLANT FSAR Section 7.3.3 and Table 7.3-5. F..rAR , "3.3.8'., :INSEft'I 8 3,3? B 3.3.8-7 Revision Bc INSERT B 3.3.8 SR 3.3.8.6 is the performance of the required response time verification every 18 months on a STAGGERED TEST BASIS on those functions with time limits provided in Reference
: 4. Each verification shall include at least one train such that both trains are verified at least once per 36 months. SR 3.3.8.6 is modified by a Note stating that the radiation monitor detectors are excluded from ESF RESPONSE TIME testing. The Note is necessary because of the difficulty associated with generating an appropriate radiation monitor detector input signal. Excluding the detectors is acceptable becaus.e the principles of detector operation ensure a virtually instantaneous response.
ATTACHMENT 5 PROPOSED FSAR CHANGES (for information only)
CALLAWAY -SP be isolated.
The purge and vent lines are closed on a containment isolation signal, thus minimizing the escape of any radioactivity.
The containment purge isolation signal may be initiated by manual action. b. Fuel Building Accident In the fuel building, a fuel assembly could be dropped in the transfer canal, in the fuel storage pool or in the cask loading pool. In addition to the area radiation monitors located on the wall around the fuel storage pool, portable radiation monitors capable of emitting audible alarms are located in this area during fuel-handling operations.
The doors in the fuel building are closed to maintain controlled leakage characteristics in the fuel storage pool region during operations involving irradiated fuel. Should a fuel assembly be dropped in the canal, in the cask loading pit, or in the pool and release radioactivity above a prescribed level, the radiation monitors sound an audible alarm. If one of the redundant discharge vent radiation monitors, GG-RE-27 or 28, indicates that the radioactivity in the vent discharge is greater than the prescribed levels, an alarm sounds and the auxiliary/fuel building normal exhaust is switched to the ESF emergency exhaust system to allow the spent fuel pool ventilation to exhaust through the ESF charcoal filters to remove most of the halogens prior to discharging to the atmosphere via the unit vent. The supply ventilation system servicing the spent fuel pool area is automatically shut down, thus ensuring controlled leakage to the atmosphere through charcoal adsorbers (refer to Section 9.4.2). The probability of a fuel-handling accident is very low because of the safety features, administrative controls, and design characteristics of the facility, as previously mentioned.
15.7.4.5.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.7-7 and 15A-1. In the evaluation of the fuel-handling accident, all the fission product release assumptions of Regulatory Guide 1.25 have been followed.
Table 15.7-2 provides a comparison of the design to the requirements of Regulatory Guide 1.25. The following assumptions, related to the release of fission product gases from the damaged fuel assembly, were used in the analyses:
: a. The dropped fuel assembly is assumed to be the assembly containing the peak fission product inventory.
All the fuel rods contained in the dropped assembly are assumed to be damaged. In addition, for the analyses for 15.7-11 Rev.OL-17 4/09
: b. c. d. e. f. g. h. i. j. k. I. CALLAWAY -SP the accident in the reactor building the dropped assembly is assumed to damage 20 percent of the rods of an additional assembly.
The assembly fission product inventories are based on a radial peaking factor of 1.65. The accident occurs 72 hours after shutdown, which is the earliest time fuel-handling operations can begin. Radioactive decay of the fission product inventories was taken into account during this time period. Only that fraction of the fission products which migrates from the fuel matrix to the gap and plenum regions during normal operation was assumed to be available for immediate release to the water following clad damage. The gap activity released to the fuel pool from the damaged fuel rods consists of 10 percent of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine contained in the fuel rods at the time of the accident.
The pool decontamination factor is 1.0 for noble gases. The effective pool decontamination factor is 100 for iodine. The iodine above the fuel pool is assumed to be composed of 75 percent inorganic and 25 percent organic species. The activity which escapes from the pool is assumed to be available for release to the environment in a time period of 2 hours. No credit for decay or depletion during transit to the site boundary and outer boundary of the low-population zone is assumed. No credit is taken for mixing or holdup in the fuel building atmosphere.
The filter efficiency for the ESF emergency filtration system is assumed to be 90 percent for all species of iodine. q () The fuel building is switched from the auxiliary/fuel exhaust system to the ESF emergency exhaust system within . seconds from the time the activity reaches the exhaust duct. The activity released before completion of the switchover is assumed to be discharged directly to the environment with no credit for filtration or dilution.
Even if fuel building ventilation isolation does not occur automatically, the calculated doses will be less than #tese-reported in Table 15.7-8 for the bounding case, inside the Response time testing ef tI Ie fuel building ventilation isolation functi0rl#. -r-:: J .-/ r, ec"'f'-15.7-12 f.er I J t11ca(1lJn
*fcr+Ae. Rev.OL-17 4/09 I. II. III. CALLAWAY -SP TABLE 15.7-7 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL-HANDLING ACCIDENT In Fuel Building Source Data a. Core power level, MWt 3,636 b. Radial peaking factor 1.65 c. Decay time, hours 72 d. Number of fuel assemblies 1.0 affected e. Fraction of fission product gases contained in the gap region of the fuel assembly Per R.G. 1.25 Atmospheric Dispersion Factors See Table 15A-2 Activity Release Data a. Percent of affected fuel assemblies gap activity released 100 b. Pool decontamination factors 1. Iodine 100 2. Noble gas 1 c. Filter efficiency, o until isolation percent 90 thereafter
: d. Building mixing volumes assumed, percent of total volume 0 e. HVAC exhaust rate, 20,000 until isolation cfm 9,000 thereafter
: f. Building isolation time, 7 sec g. Activity release period, hrs 2 90 In Reactor Building 3,636 1.65 72 1.2 Per R.G. 1.25 See Table 15A-2 100 100 1 0 0 Activity completely released over 2 hours. 2 hours 2 Rev.OL-17 4/09 CALLAWAY -SP TABLE 15.7-8 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT In Fuel Building Exclusion Area Boundary (0-2 hr) Thyroid Whole-body Low Population Zone Outer Boundary (duration)
Thyroid Whole-body In Reactor Building Exclusion Area Boundary (0-2 hr) Thyroid Whole-body Low Population Zone Outer Boundary (duration)
Thyroid Whole-body Doses (rem) -6.59 to .tfJ .f).234 0 .. ;235
(), f:rHJ ..Q.9234 /).0:23S" 61.7 0.359 6.17 0.0359 Rev.OL-17 4/09 CALLAWAY -SP TABLE 16.3-2 (Sheet 3) INITIATING SIGNAL AND FUNCTION b. Start Turbine-Driven Auxiliary Feedwater Pump c. Feedwater Isolation
: 10. Loss-of-Offsite Power Start Turbine-Driven Auxiliary Feedwater Pump 11. Trip of All Main Feedwater Pumps Start Motor-Driven Auxiliary Feedwater Pumps 12. Auxiliary Feedwater Pump Suction Low Transfer to Essential Service Water 13. RWST Level-Low-Low Coincident with Safety Injection Automatic Switchover to Containment Sump 14. Loss of Power a. 4 kV Bus Undervoltage-Loss of Voltage b. 4 kV Bus Undervoltage-Grid Degraded Voltage 15. Phase "A" Isolation
: a. b. Control Room Isolation Containment Purge Isolation
: 16. Control Room High Gaseous Activity Control Room Isolation RESPONSE TIME IN SECONDS 60(8)(17) 2(5),(8) N.A. N.A. 2(5) TABLE NOTATIONS (1) Signal actuation, diesel generator starting, and sequencer loading delays included.
Valve stroke times and spin-up times for pumps and fans included, as applicable.
(2) Diesel generator starting delay not included.
Offsite power available.
Signal actuation, sequencer loading, and pump spin-up delays included.
(3) Signal actuation, diesel generator starting and sequencer loading delays included.
RHR pumps not included.
Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is included.
Rev.OL-17k 4/10 INSERT FSAR 1 INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 17. Fuel Building Ventilation Exhaust High Gaseous Activity Emergency Exhaust System in the FBVIS Mode 90(19)
CALLAWAY -SP TABLE 16.3-2 (Sheet 5) of non-emergency AC and the loss of normal feedwater accident analyses, initiation of AFW flow is assumed delayed for 90 seconds following reactor trip on a low-low steam generator water level signal. (17) Response times noted above include the transmitters, 7300 process protection cabinets, solid state protection cabinets, and actuation devices only. For the feedline break accident analysis, initiation of AFW flow is assumed delayed for 90 seconds following reactor trip on low-low steam generator water level signal. (18) The response time for the reactor trip breakers to open and the gripper release time are satisfied by measurement and included in the response time for each required reactor trip functon. Rev.OL-17k 4/10 INSERT FSAR 2 (19) The radiation monitor detector is excluded from response time testing. The stated response time accounts for the elapsed time between introduction of a count rate from the detector corresponding to the actuation setpoint and repositioning of the components necessary to place the Emergency Exhaust System in the FBVIS mode of operation.}}

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