GNRO-2018/00048, Response to Request for Additional Information License Amendment Request, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6
| ML18284A041 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/10/2018 |
| From: | Emily Larson Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GNRO-2018/00048, NEI 99-01, Rev 6 | |
| Download: ML18284A041 (633) | |
Text
{{#Wiki_filter:A* ~Entergy Entergy Operations, Inc. P. o. Box 756 Port Gibson, MS 39150 Eric Larson Vice President, Operations Grand Gulf Nuclear Station Tel. (601) 437-7500 10 CFR 50.90 GNR0~2018/00048 October 10, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Response to Request for Additional Information License Amendment Request, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6 Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29 /
REFERENCES:
- 1. Entergy Operations, Inc. (EOI) letter to U. S. Nuclear Regulatory Commission (NRG), "License Amendment Request, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6," dated April 27, 2018 (ML18117A514)
- 2. NRG email to EOI, "Final Request for Additional Information - Emergency Action Level Scheme Change (L-2018-LLA-0116)," dated August 30, 2018 (ML18250A304). ,
Dea.r Sir or Madam, In Reference 1, Entergy Operations, Inc. (EOI) requested an amendment to Facility Op~rating License No. NPF-29 for Grand Gulf Nuclear Station, Unit 1 (GGNS). The proposed change revises the Emergency Plan for GGNS to adopt the revised Emergency Action Level (EAL) scheme described in Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the Nuclear Regulatory Commission* (NRG). In Reference 2, the NRG requested additional information to complete its review of the proposed license amendment. In response, EOI is providing the requested information in the Enclosure and associated Attachments to this letter.
GNR02018/00048 Page 2 of 2 EOI has reviewed the information supporting a finding of no significant hazards consideration that was previowsly provided to the NRC in Reference 1. The-information provided ir,i this submittal does not* affect the basis for concluding that the proposed license amendment does not involve a significant hazards consideration. There are no regulatory commitm~nts contained within this letter. Should you have any questions or require additional information, please contact Douglas Neve at 601-437-2103. . I declare under penalty of perjury, the foregoing is true and correct. Executed on October 10, 2018 Sincerely, cc::it---- Eric Larson EL/dre
Enclosure:
Response to Request for Additional Information License Amendment Request, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6 Attachment 1: GGNS Calculation XC-QID17-17001, Radiological Effluent EAL Threshold Values (EP CALC-GGNS-1701) Attachment 2: GGNS EAL Basis Document Attachment 3: GGNS EAL Basis Document Ma~kup cc: with Attachment U.$. Nuclear Regulatory Commis.sion ATTN: Mrs. Lisa Regner N RR/DOLR/LPL4 Mail Stop OWFN/ 11 F1 11555 Rockville Pike Rockville, MD 20852-2378 cc: without Attachment Mr. Kriss-Kennedy. 1 Regional Administrator, Region IV U.S. Nuclear Regula~ory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRG Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150
GNR0-2018/00048 Page 1 of _19 GNR0-2018/00048, ENCLOSURE Response to Request for Additional Information License Amendment Request Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6 Attachment 1 ~GNS Calculation XC-QID17-17001, Radiological Effluent EAL Threshold Values (EP-CALC-GGNS-1701)' Attachment 2 GGNS EAL Basis Document Attachment 3 GGNS EAL Basis Document Markup
GNR0-2018/00048 Enclosure 1 . Page 2 of 19 Response to Request for Additional .Information License Amendment Request Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6
GNR0-2018/00048 Page 3 of 19 The format for the RAI responses below is as follows: The Request for Additional Information (RAI) is provided verbatim as received from the Nuclear Regulatory Commission (NRC). This is followed by the Grand Gulf Nuclear Station (GGNS) RAI response to the individual question.
RAI 1
It appears that some text is missing in Enclosure 1 of the License Amendment Request (LAR) between pages 6 and 7. Please provide the missing text, or clarify intent. RAI 1 RESPONSE GNRO 2018/00008 Enclosure 1 Pages 6 and 7 text is replaced as follows:
- 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed changes to the GGNS EALs do not involve any physical changes to plant equipment or systems and do not alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors and do not alter design assumptions, plant configuration, or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems or components (SSCs) to perform intended safety functions in mitigating the consequences, of an initiating event within the assumed acceptance limits. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident
- previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
GNR0-2018/00048 Page 4 of 19 Response: No No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. Because EALs are not accident initiators and no physical changes are made to the plant, no new causal mechanisms are introduced. Therefore, the changes do not create the possibility of a new ordifferent kind of accident from an accident previously evaluated.
- 3. Do the proposed changes involve a significant reduction in a margin of safety?
Response: No Margin of safety is associated with the ability of the fission
*product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not impact operation of the plant and no accident analyses are affected by the proposed changes.
The changes do not affect the Technical Specifications or the method ofoperating the plant. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The .proposed changes will not result in plant operation in a configuration outside the design* basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the changes do not involve a significant reduction in a margin of safety. Based upon the reasoning presented above, Entergy
.concludes that the requested change involves no significant hazards consideration, as set forth in 10 CFR 50.92(c),
"Issuance of Amendment."
4.4 Conclusions In conclusion, based on the considerations discussed
- GNR0-2018/00048 Page 5 of 19 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) the issuance of the amendment will not be inimical to the common defense_and security or to the health and safety*of the public. (. *
5.0 ENVIRONMENTAL CONSIDERATION
The proposed changes are applicable to emergency planning requirements involving the proposed adoption of the NRG-endorsed EAL guidance as described in NEI 99-01, Revision 6, and do not reduce the capability to meet the emergency planning standards of 10 CFR 50.47(b) and the* requirements of10 CFR 50, Appendix E.* The proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or . significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet 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 changes. 6~ REFERENCES
- 1. NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels" February 2008 (ML080450149)
- 2. NRC letter "U.S. Nuclear Regulatory Commission Review and Endorsement of NE/ 99-01, Revision 6, Dated November, 2012 (TAC No. 092368), "March 28, 2013 (ML12346A463)
- 3. Entergy letter dated December 1, 2011, "Proposed Emergency Action Levels Using NE/ 99-01, Revision 5 Scheme, "-(ML12244A351). (TAC NO. ME7540) )
- 4. NRC Safety Evaluation dated October 10, 2012, "Emergency Action Level Scheme Upgrade Based on Nuclear Energy Institute 99-01, Revision 5 Entergy Operations, Inc. Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416," (TAC Nos. ME7540)
GNR0-2018/00048 Page 6 of 19
- 5. NRC letter "Callaway Plant, Unit 1 - Issuance of Amendment Re: Upgrade to Emergency Action Level Scheme (GAG No. MF4945)," October 7, 2015 (ML15251A493)
- 6. NRC letter "Fermi 2 - Issuance of Amendment to Revise the Emergency Action Level Scheme for the FERMI 2 Emergency Plan (TAC No. MF5048)," September 29, 2015 (ML15233A084)
- 7. NRC letter "South Texas Project, Units 1 and 2 - Re:
Upgrade to Emergency Action Level Scheme (TAC Nos.'\ MF4195 andMF4196)," August 20, 2015 (ML14164A341) RAl2 Concerning Table A-1, "Effluent Monitor Classification Thresholds," please address the following:
- a. For EAL AU1 .1, explain how the proposed threshold values for five different effluent flow paths with different flow rates all* have the same threshold value, since it appears that different flow* rates would require different alarm setpoints.
- b. For EALs AS1 .1 and AG1 .1, explain why the threshold values have significantly changed from the currently approved EAL scheme. This explanation. should include the change from a single threshold value
- for all gase.ous effluent flow paths to separate values for the gaseous effluent flow paths, as well as the reason for the magnitude of the change. ,
RAI 2 RESPONSE
- a. The AU1 .1 gaseous effluent EAL thresholds are based on the UFSAR Table 11.3-9 expected annual noble gas effluent release totals fo'r normal coolant (no c9re damage). The same normal coolant activity source term fractions, based on the total estimated annual activity -~
released; are applied to each release point. The thresholds in counts per minute (cpm) or microcuries per cubic centimeter (µCi/cc) are different for each release point due to the different effluent flow rates as documented in XC-01017-17001, Radiological Effluent EAL Threshold Values (EP-CALC-GGNS-1701 ). The GGNS effluent parameter display system converts monitor readings into release rate
GNR0-2018/00048 Enclosure 1 ' Page 7 of 1~ values of curies per second (Ci/sec), which will be equivalent for all release points for a given 10 CFR 20 annual exposure limit basis.
- b. Current AG1 .1 and AS1 .1 EAL thresholds were developed on a site specific dose model (Dosecalc Program) that is currently in service, process reduction factors for which are built in to the various accident .
isotopic mix spectrums available to the user. Releases filtered exch.,1sively through the Standby Gas Treatment System are subject to additional reduction of iodine in the Dosecalc Program, but other physical release pathways-do not have individual process reduction factors built in. The proposed NEI 99-01 based AG1 .1, AS1 .1 and AA 1.1 thresholds are developed using a GGNS site specific dose assessment model which is intended to be used when the proposed EAL scheme is implemented at GGNS. The dose model applies different process reduction factors for the fuel clad accident source term mix depending on the release pathway, which alters the composition of th,e activity released from each pathway. This method of process reduction is based on NUREG-1940. The threshold development methods between the current and the proposed EALs are not the same. Other differences Dosecalc and the intended model process reduction factors, such as X/Q and source term, account for further differences in magnitude. RAl3 For EALs AA2.3, AS2.1, and AG2.1, provide additional detail for the basis for
- rounding the threshold values upwards by approximately 10 inches, as this could result in an early and/or unnecessary declarations for a site area emergency or general emergency classification.
RAI 3 RESPONSE The original submittal contained two typographical errors. Level 2 is 193 ft 2 1/8 inches NOT 192 ft 2 1/8 inches. Level 3 is 183 ft 2 1/8 inches NOT 182 ft 2 1/8 inches. Changes were made to the applicable pages of the Grand Gulf Nuclear Station EAL Basis Document to correct the typographical error and show that the instrument for spent fuel [pool level is not located in the Control Room. Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections. I
GNR0-2018/00048 Page 8 of 19 RAl4 For EALs CS1 .3 and CG1 .2, the Basis discussion for the proposed threshold 1 value for the containment radiation monitors provides that the detectors are in a position to monitor the containment radiation environment above the refueling cavity elevation. Additional justification is provided in the EAL Comparison Matrix, which states that the threshold value "is indicative of likely core uncovery in the refueling zone." Provide additional detail supporting the threshold value for the proposed containment radiation monitors .
. RAI 4 RESPONSE Upon review of the calculations and the RAI question, Entergy has determined that using the proposed threshold values would have had the potential to cause excessive confusion to the operator making the EAL determination. Therefore, Entergy will apply currently in use NRC approved NEI 99-01 Revision 5 EAL values to the proposed NEI 99-01 Revision 6 scheme to ensure that the EAL are accurately classified. These threshold values were approved via NRC Safety Evaluation dated October 10, 2012, "Emergency Action Level Scheme Upgrade Based on Nuclear Energy Institute 99-01, Revision 5 Entergy Operations, Inc. Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416," (ML12244A351). The NEI 99-01 Revision 5 methodology for these EALs is the same as the NEI 99-01 Revision 6 methodology. Therefore, applying the NEI 99-01 Revision 5 EAL threshold values to NEI 99-01 Revision 6 does not represent a selective application of
. an EAL threshold from one EAL scheme to another.
RAIS The proposed ~AL CU3.1 contains the condition "... due to the loss of decay heat removal capability," which is not consistent with NEI 99-01, Revision 6. This could result fn potential misclassification for an event that causes reactor coolant system (fKCS) temperature to rise above 200 degrees Fahrenheit (°F) when decay heat removal capability has not been lost. Provide additional detail for adding the condition, " ... due to the loss of decay heat removal capability," to the EAL CU3.1 threshold value, or revise accordingly.
GNR0-2018/00048 Page 9 of 19 RAI 5 RESPONSE Entergy agrees with the noted concern and, subsequently, the subject phrase has been removed from the CU3.1 EAL. The EAL will now state: 11 "UNPLANNED rise in RCS temperature to> 200 °F. In addition, the first paragraph of the GGNS basis for CU3.1 is changed from "Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200 °F) (ref. 3). In. the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification is based on the concurrent loss of reactor vessel level indications_ per EAL CU3.2" to "Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200 °F) (ref. 3). In the absence of reliable RCS temperature indication, classification is based on t~e concurrent loss of reactor vessel level indications per EAL CU3.2." Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL
/
Basis Documents (clean and markup) to make these corrections. RAl6 The proposed EAL CA3.1 Basis discussion (1st paragraph) contains the condition, " ... caused by the loss of decay heat removal capability," which is not consistent with NEI 99-01, Revision 6. This could result in potential misclassification for an event other than a loss decay heat removal capability that leads to an unplanned RCS pressure increase. Provide additional detail for the proposed Basis wording, or revise accordingly. RAI 6 Response Entergy agrees with the noted concern and, subsequently, the subject phrase has been removed from the 1st paragraph of the EAL CA3.1 Basis Discussion. The. paragraph will now state: "In the absence of reliable RCS temperature indication, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4."
GNR0-2018/00048 Page 10 of 19 Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections .
. RAl7 For EALs CU5.1 and SU7.1 explain how the INFORM Notification System (INFORM) can be used as a State and local agency communication method.
This response should explain whether or not INFORM is indep*endent of the provided telephone systems and if INFORM supports two way communications. RAI 7 Response Although INFORM is independent of the other telephone systems provided in the associated tables for these EALs, it does not support two-way communication and is, therefore, removed from Tables C-5 and S-4 in EALs CU5.1 and SU7.1. Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections. RAIS The proposed RCS barrier (RCB) 2 on the fission product matrix does not include the high pressure core spray (HPCS) system. The guidance states that the list of systems should also include high pressure coolant injection [high pressure core spray], since a rupture of the HPCS, if not isolated, could rapidly depressurize the reactor pressure vessel. Please justify not including the HPCS as a threshold value for the proposed RCB2. RAI 8 Response Review of the RAI determined that the original submittal needs to be revised to include the "HPCS line break unisolable from the reactor coolant system EAL. Additionally, clarification has been added to th~ basis section of "Even though the High Pressure Core Spray (HPCS) injects into the RCS, it is included in this EAL due to the potential for an inter-system loss of coolant back flowing from the discharge lines (via failed isolation valves and check valves) and out
GNR0-2018/00048 Page 11 of 19 through a break in the piping. A HPCS failure that does not result in back flow a of RCS an_d out through break should not be considered as meeting the EAL threshold." Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections. RAl9 The proposed threshold values for fission product barrier degradation based on containment radiation monitors do not appear valid. Considering that the Fuel Clad Barrier (FCB) Loss threshold value should correspond to 2% to 5% clad damage, and the Containment Barrier (CNB) Potential Loss threshold value should be 20%, as provided by NEI 99-01, Revision 6, it would be reasonable for the radiation values to be-different by a factor of 4 to 10. However, the value for the CNB Loss radiation monitor reading is 17 .5 times higher than the FCB Loss radiation monitor reading. Additionally, the NRC staff could not determine why the threshold value for the FCB3 Loss is significantly lower than that for River Bend Station (RBS), which is a lower powered Boiling Water Reactor Type 6 (BWR..,6) that also has a Mark 3 Containment (400 R/hr for GGNS and 3000 R/hr for RBS), while the CNB threshold values were much closer (7000 R/hr for GGNS and 12000 R/hr for RBS). Please verify that the radiation monitor threshold values for a FCB Loss are based on a loss of the RCS with between approximately 2% and 5% clad damage and that the radiation monitor threshold values for a CNB Potential Loss are based on approximately 20% clad damage. RAI 9 Response In a_ccordance with a follow-up clarification call held with NRC, the purpose of the RAI is not to compare the GGNS and RBS calculation results but to ensure the basis for both sites' EAL thresholds is correct. The NEI 99-01, Revision 6, developer notes direct that "the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 uCilgm dose equivalent 1-131, into the primary containment atmosp~ere."The associated NEI basis states in part "reactor coo/ant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage."
GNR0-2018/00048 Page 12 of 19 The 2% - 5% fuel clad damage in the NEI basis for PWRs and BWRs orig.inated in NUMARC/NESP.-007 Rev 2 and could be confirmed / using typical enrichment values and NUREG-1228 related source. term/partitioning assumptions. Currently, most units operate with higher enrichment than what was common in the late 1980s and have been approved for power uprates. Additionally, newer source term and partitioning guidance such as NUREG::- 1940 are a factor. These combine such that typical reactor coolant concentrations to percent clad damage are approximately half of what was calculated using historical inputs and guidance. In a,ddition, the use of a 300 uCi/gm dose equivalent 1-131 (DEi) source term for FCB3 (NEI Fuel Clad Barrier Loss threshold 4.A) provides agreement within the EAL scheme with FCB4 (NEI Fuel Clad Barrier Loss threshold 1.A) which directly refers to a 300 uCi/gm DEi value. The calculated radiation monitor threshold value for Potential Loss of the Containment Barrier in CNB8 is based on 20% clad damage. Therefore, GGNS believes this use of 300 uCilgm dose equivalent 1-131 more accurately meets the intent of NEI 99-01, Revision 6, developer notes and should be used if there is a disagreement between 300 uCilgm dose equivalent l-131and the 2-5% fuel clad damage.
RAI 10
Explain why the Basis discussion (third paragraph) for a Potential Loss under CNB7, which* states, "cannot be maintained above," does not use the same wording as the threshold value, which states, "cannot be restored and maintained within." This difference in wording could result in an inaccurate or delayed assessment. RAI 10 Response Entergy agrees with the noted concern. The first sentence of the third paragraph of the basis for CNB7 is changed from ""the term "cannot be restored and maintained above" means the parameter value(s) is not able to be brought.within the specified limit" to '"'the term "cannot be restored and maintained within" means the parameter value(s) is not able to be brought within the specified limit" Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean a.nd markup) to make these corrections.
GNR0-2018/00048 Page 13 of 19
RAI 11
The proposed EAL HU4'.2 - Table H-1, "Fire Areas," includes the Containment Building in all_modes. This could result in an event declaration* due to the spurious actuation of a single fire alarm. The NRC staff could not determine if the Containment Fire Detection System, in combination with the Containment Ventilation System, supported the inclusion of the Containment Building as a fire area for EAL HU4.2. Provide justification that demonstrates why GGN includes-the Containment Building in the Table H-1 for all modes, or modify accordingly. RAI 11 Response EAL HU4.1 addresses the condition where a fire is reported and verified in a listed Fire Area. This verification could be from a report in the field or because multiple fire detection device alarms are received. This EAL includes a table that lists fire areas of concern, including containment. EAL HU4.2 addresses receipt of a single fire detector without a corresponding verification. Entergy proposes to make an exception in EAL HU4.2 to exclude the containment and drywell in Modes 1 and 2. Personnel safety concerns preclude entry into certain areas of containment and there are areas within containment where fire detectors are located that would be inaccessible during these modes due to elevated radiation levels. Industry experience has demonstrated that including containment in Modes 1 and 2 in EAL HU4.2 can lead to unusual event emergency classifications based on a single spurious fire alarm, requiring subsequent emergency retractions. With regard to containment and drywell fire alarms, it can reasonably be
. expected that a fire that bums for 15 minutes would produce sufficient products of combustion to cause multiple fire detection devices to alarm. This is due to the products of combustion being transported to other areas inside the conta.inmenUdrywell due to the forced flow ventilation system in operation.
Receipt of a single fire alarm would likely be due to a spurious detector actuation. There are four Containment Fan Cooler (CFC) units located in the Grand Gulf Nuclear Station Unit 1 (GGNS) containment building. Each CFC fan delivers approximately 75,000 CFM in normal mode and 37 ,800 CFM in accident mode. The four CFC units operate in accident mode when a Safety Injection Signal is present. Two or three of the four CFCs are operating in normal mode at any given time to cool the Containment in modes 1 or 2. The CFC units draw return air from the containment atmosphere ~nd discharge into a common header which discharges to multiple areas inside containment. This*
GNR0-2018/00048 Page 14 of 19 constant flow of air would draw any smoke/heat towards the cooling units past the installed detectors, thus initiating multiple detector alarms. Actuation . of more than one detector is the most reliable indication of an actual fire because of high volumetric air flow throughout the containment building. Due to construction of the intermediate floors and multiple openings in the floors, it can be expected that smoke/heat would migrate throughout containment in a very short period and that 2 or more detectors would alarm. Basing emergency classifications on receiving more than one detector actuation is therefore the most reliable indication of a valid alarm and accurately meets the Initiating Condition of HU4, "FIRE potentially degrading the level of safety of the plant." The drywell cooling system consists of recirculating fan-coil units and the associated dampers, ducting, and cont~ols required to maintain the design drywell temperature and relative humidity. Each fan-coil unit consists of two full-capacity fans in parallel and two full-capacity cooling coils in series. Six fan-coil units with a capacity of 12,000 cfm per fan are provided to distribt.1te cooling air effectively and with minimum ductwork. Normally, one fan and one coil of each fan-coil unit operate, and the other fan and coil ,are on standby. Additionally, the drywell cooling system incorporates two recirculation fans with a capacity of 6,000 cfm per fan and associated controls and ducting which transfer air from the upper elevation to the lower elevation. These fans alleviate heat stratification in the drywall. Normally, both fans operate simultaneously. Actuation of more than one smoke detector is the most reliable indication of an actual fire because of high volumetric air flow throughout the drywell. Due to construction of the intermediate floors and multiple openings in the floors, it can be expected that smoke/heat would migrate throughout the drywell in a very short period and that 2 or more detectors would alarm. Basing emergency classifications on receiving more than one detector actuation is therefore the most reliable indication of a valid alarm and accurately meets the Initiating Condition of HU4, "FIRE potentially degrading the level of _safety of the plant." With consideration to the above discussion, Note 11 is added to EAL HU4.2 as follows:
"During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the containment or drywell."
The following information is added to the Basis for HU4.2:
GNR0-2018/00048 Page 15 of 19 This EAL is not applicable for the containment or drywell in Modes 1 and 2. The air flow design and TS requirements for operation of Containment Fan Coolers and the drywell cooling system are such that multiple detectors would be expected to alarm for a fire in the containment or drywell. A fire in the containmenror drywell in these modes would therefore be classified under EAL HU4.1. Verification of a single containment or drywell fire alarm that is likely to be spurious does not warrant the potential elevated exposure risks and industrial safety risks associated.with an emergency entry of containment or drywell in modes 1 and 2. Therefore, GGNS proposes to make EAL HU4.2 applicable to a single fire alarm in containment or dryweU in Modes 3, 4and 5. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003, it is reasonable to conclude that the changes proposed to EA~ HU4.2 would be considered a deviation from the formally endorsed guidance of NEI 99-01 Revision 6. The structure ofthe proposed deviation for HU4 IC/EAL is modelled after Seabrook Station's adoption of NEI 99-01 Revision 6 EALs containing a similar exception, which was approved by the NRG with Amendment 152 to the Seabrook Station Facility Operating License No. NPF-86 on February 10, 2017 (ADAMS Accession No. ML16358A411 ). Proposed Emergency Preparedness Frequently Asked Question 2018-003 (ADAMS Accession No. ML18081A309) was also used as a basis for this deviation. Based on the information above, Entergy considers the proposed revision to be an acceptable deviation from the generic NEI 99-01, Revision 6, guidance. This deviation is consistent with proposed Emergency Plan (EP) Frequently Asked Question (FAQ) 2018-03 (ML18081A309). Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections.
RAI 12
The proposed EAL SU4.1 threshold value is based on the Offgas Pretreatment Radiation Monitor High-High Alarm, while the currently approved EAL scheme uses a table that includes various radiation monitor readings, which correspond to various flow rates.
. -- I
GNR0-2018/00048 Page 16- of 19 The NRC staff could not determine if a value that was approximately equal to the technical specification allowable limits could be assessed with the proposed threshold value. Provide justification that supports using the Offgas - Pretreatment Radiation Monitor High-High Alarm as a threshold value for SU4.1. This Justification should include a discussion of the difference between the currently approved EAL scheme (EAL SU9.1) and the proposed EAL SU4.1 RAI 12 Response During development of the NEI 99-01 Revision 6 EALs for GGNS it was determined that it was more effective and convenient to the Operator to use the Offgas Pretreatment Radiation Monitor High-High Alarm as a threshold value for SU4.1. GGNS NEI 99-01 Revision 5 .EALs used manual comparison of flow rates to radiation monitor readings to determine when Technical Specification {TS) 3.7.5 value of 380 millicuries per second release rate has been met or exceeded. Revision 6 EAL will use an existing Alarm (Offgas Pretreatment Radiation Monitor High-High) that is driven by a computer calculation to determine the Initiating Condition (IC) is met. The purpose of the Offgas Pretreatment Radiation Monitor High-High Alarm is to come into alarm when the Technical Specification 3.7.5 value of 380 millicuries per second release rate has been met or exceeded. This is the initiating condition for SU4.1. By using this alarm, it allows the Operators to diagnose entry into SU4.1 quickly rather than having to review flowrates and radiation monitor readings. The use of flowrates and radiation monitor readings is a viable contingency action that will be maintained in the GGNS procedures and is added to the basis document information. Enclosure Attachments 2 and 3 contain the Grand Gulf Nuclear Station EAL Basis Documents (clean and markup) to make these corrections.
GNR0-2018/00048 Page 17 of 19 GNR0-2018/00048, ENCLOSURE ATTACHMENT 1 GGNS Calculation XC-QID17-17001, Radiological Effluen~ EAL Threshold Values (EP CALC-GGNS-1701)
O AN0-1 O AN0-2 (81 GGNS DIP-2 O IP-3 DPLP OJAF D PNPS . ORBS ovv 0W3 O NP-GGNS-3 O NP-RBS-3 1 2 CALCULATION <> EC# 73157 < >Page 1 of 72 COVER PAGE (3) Design Basis Cale. IZJ YES ONO (4) IZJ CALCULATION DEC Markup 5 5 ( > Calculation No: XC*Q1017*17001 < J Revision: 0
Title:
Radiological Effluent EAL Threshold Values (EP-CALC* 5 llJ < >. Editorial GGNS-1701) DYES ~NO 10 ll:IJ .* System(s}: 017 t ' Review Org (Department}: 11 12 < > Safety Class: < >
- Component/Equipment/Structure Type/Number:
D Safety I Quality Related IZJ*Augmented Quality Program D Non~Safety Related 13
< > Document Type: JOS.02 14
< > Kevwords (Descriptionfropical ' Codes}: EAL REVIEWS 15 15 17 Name/Signature/Date Name/Signature/Date > Name/Si n ~/
-;~e F1 Scott McCain, OSSI, Inc. Robert W. Fuller f..d!;t/Q r4. '2.*l.f>- Je Brandon Tavlor,,,,, / ~
- ~
Responsible Engineer ~ Design Verifier Supervisor/Approval D Reviewer D Comments Attached D Gomments Attached
~Enterrn bJ@
- i' Grand Gulf
-Nuclea r Station
~~~-,~~,;;;;;-~7-:::,,.:.,,~.1q;~%:;__,~-'.,'.;_:_c;;;;,;~~ ( G G NS) Radiolo gical Effluen t EAL Thresh old Values EP-CALC-GGNS-1701 Revision O OSSI Author: Scott McCain 02/07/18
GGNS EAL Technica*I Bases Calculations - Ax1 Effluent Series Table of Contents
- 1. PURPOSE ............................................................................................................................... 3
- 2. DEVELOPMENT METHODOLOGY AND BASES .................................................................. 3 2.1. Threshold Limits ........................................................................................... :... :************* 3 2.2. Effluent Release Points .................................................................................................. 6 f
2.3. Source Term **********************************************************************:******************************************* 8 2.4. Effluent Flow................................................................................................................. 10 2.5. Release Duration .......................................................................................................... 11 2.6. Meteorology ************************************************************************************************************:**** 11
- 3. DESIGN INPUTS .................................................................................................................. 13 3.1. General Constants and Conversion Factors ................................................................ 13 3.2. Liquid ~ffluent .............................................. .:.............................................................. 13 3.3. Gaseous Effluent ............*.............................................................................................. 13
- 4. CALCULATIONS ................................................................................................................... 18 4.1. AU1 .1 Liquid Release********************************************;***************************************************** 18 4.2. AU1.1 Gaseous Release .............................................................................................. 19 4.3. AA 1.1, AS1 .1 and AG1 .1 Gaseous Release ................................................................ 19
- 5. CONCLUSIONS ...... *.............................................................................................................. 21 5.1. Effluent Monitor Reading Results in CPM (All Calculated Values) ............................... 21 5.2. Effluent Monitor Reading Results in CPM (For Applicable EAL Thresholds) ............... 22 .\.
5.3. Gaseous Effluent Monitor Reading Results in Ci/sec ................................................... 22
- 6. REFERENCES ............................................................................ ; ......................................... 23 ATTACHMENTS , AU1 .1 Liquid Effluent EAL Calculations ............................................................... 25 , AU 1.1 Gaseous Effluent EAL Calculations .......................................................... 26 , SBGT B Canberra Monitor Correlation ................................................................ 28 , AA 1.1, AS1 .1 and AG 1.1 URI Calculations ......................................................... 29 EP-CALC-GGNS-1701 Page 2 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series
- 1. PURPOSE The Grand Gulf Nuclear Station (GGNS) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Revision.6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of the gaseous and liquid radiological effluent EAL values developed in accordance with the guidance in NEI 99-01 Revision 6.
Documentation of the assumptions, calculations and results are provided for the GGNS Ax1 series EAL effluent monitor values associated the NEI 99-01 Revision 6 EALs listed below.
- NEI EAL AU 1.1 (gaseous and liquid)
- NEI EAL AA 1.1 (gaseous and liquid)
- NEI EAL AS1 .1 (gaseous)
- NEI EAL AG1 .1 (gaseous)
- 2. DEVELOPMENT METHODOLOGY AND BASES 2.1. Threshold Limits 2.1.1. AU1 .1 Liquid Threshold Limits Guidance Criteria The AU1 Initiating Condition (IC) addresses a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer.
GGNS Bases ODCM Section 6.11.1 states that the limits for the concentration of radioactive liquid effluents released from the site to unrestricted areas are as follows:
- Ten (10) times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases
- 2.0E-04 µCi/ml total activity for dissolved or entrained noble gases The site specific AU1 .1 liquid effluent EAL threshold values will equate to 2 times the ODCM limit. Refer to Section 4.1 for the threshold calculation related to this limit.
EP-CALC-GGNS-1701 Page 3 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 2.1.2. AU 1.1 Gaseous Threshold Limits Guidance Criteria The AU1 Initiating Condition (IC) addresses a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer. GGNS Bases
- ODCM Section 6.11.4 states that the limits for the radioactive gaseous effluents released from the site at or beyond the site boundary are as follows:
- Less than or equal to 500 mrem/yr to the total body (Noble Gases)
- Less than or equal to 3000 mrem/yr to the skin (Noble Gases)
- Less than or equal to 1500 mrem/yr to any organ (1-131, 1-133, tritium and radioactive materials in particulate form with half-lives> 8 days)
ODCM gaseous setpoint calculations are based on the noble gas limits. Organ dose includes inhalation, ingestion and deposition pathways and are applied in unrestricted area site boundary gaseous effluent dose calculations used in the Annual Radioactive Effluent Release Report. Ingestion pathway bases are not compatible or directly comparable with short term event considerations, and are not a significant contribution to the total dose (total body or skin dose limits from noble gas are the major exposure pathway). Thus, the organ dose limit is not applicable for EAL threshold determination. The site specific AU1 .1 gaseous effluent EAL threshold values will equate to 2 times the ODCM limit for the lesser of the total body or skin exposure pathways. Refer to Section 4.2 for the threshold calculation related to this limit. 2.1.3. AA 1.1 Liquid Threshold Limits Guidance Criteria The AA 1 Initiating Condition (IC) addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE. This is based on values at 1% of the EPA Protective Action Guides (PAGs). Per NEI 99-01, the effluent monitor readings should correspond to the above dose limits at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. GGNS Bases The liquid effluent limits are based on the water concentration values* given in 10 CFR 20 Appendix B Table 2 Column 2 (see Section 2.1.1 above). The 10 CFR 20 values are equivalent to the radionuclide concentrations which, if ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 mrem). The EPA PAGs *are based on a TEDE dose from immersion, inhalation and deposition. EP-CALC-GG NS-1701 Page 4 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Serjes The 10 CFR 20 limits and the EPA limits do not represent the same type of exposure and thus cannot be compared on a one to one basis. Thus, the site specific EALs will not contain an AA 1.1 liquid effluent monitor threshold value that equates to 1% of the EPA PAG. However, EALs AA 1.3 (liquid effluent sample analysis) and AA 1.4 (field survey results) will remain applicable for liquid effluent releases that exceed their respective thresholds. Since EALs AA 1.3 and AA 1.4 are not associated with a calculation no further reference is made to those EALs. 2.1.4. AA 1.1 Gaseous Threshold Limits Guidance Criteria The AA 1 IC addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEOE or 50 mrem thyroid COE. Per NEI 99-01, the effluent monitor readings are based on values at 1% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. GGNS Bases r The gaseous effluent limits for AA 1.1 are based on values that equate to an offsite dose greater than 1O mrem TEOE or 50 mrem COE thyroid, which are 1% of the EPA PAGs. 2.1.5. AS1 .1 Gaseous Threshold Limits Guidance Criteria The AS 1 IC addresses a release of radioactivity resulting in offsite dose greater than 100 mrem TEOE or 500 mrem thyroid COE. I This is based on values at 10% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for, one hour of exposure. GGNS Bases The gaseous effluent limits for AS1 .1 are based on values that equate to an offsite dose greater than 100 mrem TEOE or 500 mrem COE thyroid, which are 10% of the EPA PA Gs. 2.1.6. AG1 .1 Gaseous Threshold Limits Guidance Criteria The AG 1 IC addresses a release of radioactivity resulting in offsite dose greater than 1,000 mrem TEOE or 5,000 mrem thyroid COE. This is based on values at 100% of the EPA Protective Action Guides (PAGs) at the "site-specific,dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure.* EP-CALC-GGNs.:1701 Page 5 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series GGNS Bases The gaseous effluent limits for AG 1.1 are based on values that equate to an offsite dose greater than 1,000 mrem TEDE or 5,000 mrem COE thyroid, which are 100% of the EPA PA Gs. 2.2. Effluent Release Points Note - All effluent release points assume a background reading of zero to conservatively account for all modes of operation applicable to the EALs. 2.2.1. Liquid Release Points Guidance Criteria Per NEI 99-01, the AU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (NEI AU1 EAL #1) and planned batch releases from non-continuous release pathways (NEI AU1 EAL #2). Per NEI 99-01, the AA 1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The "site-specific monitor list and threshold values" should be determined with consideration of the selection of the appropriate installed gaseous and liquid effluent monitors. GGNS Bases The single liquid effluent monitor at GGNS (ODCM Figure 1.3-1 and Table 6.3.9-1) is the Liquid Radwaste Effluent Line. 2.2.2. Gaseous Release Points Guidance Criteria Per NEI 99-01, the AU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (NEI AU1 EAL #1) and planned batch releases from non-continuous release pathways (NEI AU1 EAL #2). Per NEI 99-01, the AA 1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. EP-CALC-GGNS-1701 Page 6 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series Per NEI 99-01, the AS1 and AG1 ICs address monitored and un-monitored releases of gaseous radioactivity. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. The "site-specific monitor list and threshold values" should include the effluent monitors described in emergency plan and emergency dose assessment procedures. GGNS Bases There are seven GGNs release points to the environment. The Offgas Pre- and Post-Treatment Monitors are upstream of the Radwaste Building Vent monitor and thus are not used as separate EAL gaseous effluent threshold values. The seven gaseous effluent monitors at GGNS (ODCM Figure 2.5-1 and Table 6.3.10-1) are as follows:
- SBGT Exhaust A
- SBGT Exhaust B
- Containment Vent
- Fuel Handing Area Vent
- Radwaste Building Vent
- Turbine Building Vent
- Turbine Building occa_sional release point In modes 1, 2 and 3, the south-e.ast most smoke hatch of the turbine building may be used as an occasional release point provided that the proper portable monitoring equipment is used. During Modes 4 & 5 up to four roof hatches may be used and release rates estimated based on calculated flow rates and measured activity. Since releases from this point are infrequent, the temporary nature of the equipment, and various configurations involved, equipment substitution may occur with approval of Radiation Protection Manager and Chemistry Manager (UFSAR Section 9.4.4.2, ODCM Figure 2.5-1, 06-HP-1017-V-0001 Section 5).
Based upon the variations in portable equipment and set-up, and infrequent occurrence, the Turbine Building occasional release point pathway does not meet the NEI 99-01 criteria as a normally occurring continuous release or a planned batch release point and thus is not used as an effluent monitor EAL threshold. EP-CALC-GGNS-1701 Page 7 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 2.3.
- Source Term 2.3.1. AU1 .1 Liquid Source Term
- Guidance Criteria NEI 99-01 does not provide specific guidance for AU1 liquid source term assumptions.
GGNS Bases The AU1 .1 liquid effluent EAL threshold is based upon measured gamma emitter activity from discharge permits 2017007, 2017008 and 2017009. ,Jhe total activity of each isotope released (µCi/ml) is normalized to a representative fraction which is then adjusted to the ODCM limit (~efer to Section 2.1.1). Na-24 t - - - - - - + -8.-75E-08 - - - - - - - - t - 8.
-75E-08
- - - - + - 1.4E-02
----~
Mn-54
\I 9.53E-08 t------+------t------t------+-- 3.21 E-08 3.35E-07 4.62E-07 7.3E-02
---~
Co-60 6.17E-07 4.23E-07 1.27E-06 2.31 E-06 3.6E-01 . 2n-65 -------------+------------------- 3.04E-07 4.01 E-07 1.32E-06 2.03E-06 3.2E-01
-------------+------------------
Ag -11 Om ___ 2_.5_1_E-_0_7_ _ _ 4_.5_8E_-_0_8_ _ _ _ _ _ _ _ 2._9_7E_-_07_ _ _4_.7_E_-0_2_ Sb-125 1. 71 E-07 1. 71 E-07 2. 7E-02
-------------+------------------
Cs-134 t - - - - - - + - - - - - 3.92E-07 3.92E-07
-----t------+-----~ 6.2E-02 Cs-137 6.18E-07 6.18E-07 9. 7E-02 Totals 1.27E-06 9.89E-07 4.11 E-06 6.36E-06 1.0E+OO 2.3.2. AU1 .1 Gaseous Source Term Guidance Criteria NEI 99-01 does not provide specific guidance for AU1 gaseous source term assumptions.
GGNS Bases The AU1 .1 gaseous effluent EAL threshold is based upon UFSAR Table 11.3-9, Expected Annua11Release of Gaseous Effluents, Noble Gas (activity and fractions) for normal coolant (no core damage).
*Release 1Rate \ ,>.Noble *Gas::':
*t ?f:Jc1iJr::;.****.:~ i'./'r=t~ctiori(:rfr Ar-41 t - 7.2E+01 - - - - 9.0E-03 ----~
Kr-83m t--_o_.o_E_+_oo_--+-_o_.o_E_+_oo_~ Kr-85m t---8_.8_E_+_0_1-i---1_.1_E_-_02_~ Kr-85 t--_3_.9_E_+_02_--+-_4_:9_E_-_02_~ Kr-87 t--_6_.3_E_+_0_1-i---7_.8_E_-_03_-t Kr-88 i---9_.8_E_+_01_-+-_1_.2_E_-0_2_....... Kr-89 i--......;;6_.1;....E_+_02_-+-_7_.6_E_-o_2_....... Kr-90 i--......;;o;.....o;....E_+_oo_-+-_o_.o_E_+_oo_....... Xe-131 m i---2;.....0;....E_+_01_-+-_2_.5_E_-0_3_....... Xe-133m i--......;;.o_.o;....E_+_oo_,.-+-_o_.o_E_+_oo_~ Xe-133 Xe-135 m EP-CALC-GGNS-1701 2.2E+03 9.9E+02 2.7E-01 1.2E-01 Page 8 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series Reh:~ase Rate .Noble Gas
.(Ci/y) .. *.Fraction:
Xe-135 1.2E+03 1.SE-01 Xe-137 1.3E+03 1.6E-01 Xe-138 1.0E+03 1.2E-01 Totals 8.0E+03 1.0E+OO 2.3.3. AA1 .1. AS1 .1 and AG1 .1 Gaseous Source Terms Guidance Criteria NEI 99.:.01 specifies that the calculation of monitor readings will require use of an assumed release isotopic mix; the selected mix should be the same for ICs AA 1, AS1 and AG1. GGNS Bases The AA 1..1, AS1 .1 and AG1 .1 gaseous EAL thresholds are based upon the GGNS URI dose model results using input assumptions applicable.to the event, pathway and particular monitor. The source term used in the URI dose model is taken from NUREG-1940 Table 1.1 (URI Requirements Specification Appendix A Section A.2). The process reductions used in the URI dose model are taken from NUREG-1228 and NUREG-1465 (URI Requirements Specification Appendix A Sections A.4 and A.5). Note - HUT is hold-up time. Other than the fuel handling accident scenario, the release paths selected were chosen to represent a LOCA type event with fuel clad damage and process reductions for applicable suppression pool and bypass release pathways. URI input assumptions for the gaseous release points are as follows: Containment RCS Pool Filters Aux Bldg HUT <2 hrs SBGT Vent Env Subcooled Working HUT <2 hrs Soravs Off Release path 'K' selected to model a LOCA type event with fuel clad damage and suppression pool reduction. Containment RCS Pool Filters HUT <2 hrs 20" Cont Vent Env Subcooled Working Sprays Off Release path 'I' selected to model a LOCA type event with fuel clad damage and suppression pool reduction. Aux Bldg RW Bldg Filters RCS Radwaste Vent Env HUT <2 hrs HUT <2 hrs Workin Release path 'R' selected to model a LOCA type event with fuel clad damage and bypass reduction. EP-CALC-GGNS-1701 Page 9 of 71 Revision O
GGNS EA'L Technical Bases Calculations - Ax1 Effluent Series Turbine Building Filters RCS Turbine Building Vent Env HUT 2-24 hrs Workin Release path 'C' selected to model a LOCA type event with fuel clad damage and bypass reduction. Spent Fuel Filters Aux Building Vent Env Under Water Workin HUT 2-24 hrs Release path 'U' selected to model a spent fuel pool accident. For RCS initiated accidents, a 1 hour time after shutdown (TAS) is used for the source decay period as it is long enough for plant conditions to deteriorate for co're damage to occur and a significant release to start For the spent fuel accident, the riew fuel age option is used with a default of 80 hours for time after shutdown (TAS). 2.4. Effluent Flow 2.4.1. Effluent Liquid Discharge Flow Guidance Criteria NEI 99-01 does not provide specific guidance for effluent liquid flow assumptions. GGNS Bases Per UFSAR 11.2.1.1, the design objective of the liquid Radwaste system is to collect, process, recycle or dispose of potentially radioactive wastes produced during the operation of the plant. These wastes are grouped as floor drains, equipment drains, and chemical waste. Liquid waste collected in the equipment drain processing system is normally transferred to the condensate storage tank after' processing. Chemical wastes are sent to the floor drain collector tank for further processing or returned to the condensate storage tank. Liquid waste collected in the floor drain processing system is normally treated and released to the environment but may be recycled to the condensate storage tank. Any of these treated wastes may be discharged to the environment, providing proper dilution at the discharge basin is maintained; however, normally only processed waste from the floor drain and chemical waste subsystems will be discharged to the environment. A representative maximum discharge flow rate of 100 gpm and a minimum dilution flow rate of 3,500 gpm is used as the input for purposes of the EAL calculations (liquid release permit reports 2017007, 2017008 and 2017009). Referto Sections 3.2.3 and 3.2.4 for the input values related to this parameter. EP-CALC-GGNS-1701 Page 10 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 2.4.2. Effluent Gaseous Vent Flow Guidance Criteria NEI 99-01 does not provide specific guidance for effluent .gaseous vent flow assumptions. GGNS Bases Vent flow values for AU1 .1 are taken from 08-S-03-22. Vent flow values for AA 1.1, AS 1.1 and AG1 .1 are taken from System Flow Diagrams. Refer to Sections 3.3.3 and 3.3.4 for the input values related to the vent flow parameter. 2.5. Release Duration Guidance Criteria Per NEI 99-01, the effluent monitor readings for AS1 .1 and AG1 .1 gaseous EAL threshold values should correspond to a dose at the "site-specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure. GGNS Bases The effluent monitor readings for AA 1.1, AS1 .1 and AG1 .1 gaseous EAL threshold values are calculated for a release duration of one hour. 2.6. Meteorology Guidance Criteria The effluent monitor readings should correspond to the applicable dose limit at the "site-specific dose receptor point." The "site-specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on-site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and the procedural methodology used to determine offsite doses and protective action recommendations. This is typically the boundary of the Owner Controlled Area. Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same
. for ICs AA1, AS1 and AG1.
GGNS Bases The site specific meteorology used for the EAL calculation inputs are based upon the UFSAR and ODCM as documented below. 2.6.1. ODCM Gaseous Dispersion Factor (ODCM Table 2.2-3a} GGNS uses a ground level release model for all effluent release points (ODCM Section 2.3). EP-CALC-GGNS-1701 Page 11 of 71 Revision O
i I GGNS EAL Technical Bases Calculations - Ax1 Effluent Series The ODCM highest historical annual average X/Q at the site boundary of 4.-1 E-06 sec/m 3 is based on a wind direction from the Northeast into SW sector 'L' (045°). 2.6.2. Stability Class UFSAR Section 2.3.2.1.1.1 and Tables 2.3-130A through 130G document the predominant stability class as 'D'. Thus, a stability class of "D" is used as the URI input for purposes of the EAL calculations. 2.6.3. Wind Speed UFSAR Section 2.3.2.1.1.1, Figure 2.3-2, Figure 2.3-3 and Table 2.3-34 document the annual average wind speed of 4.4 mph. Thus, a wind speed of 4.4 mph is used as the URI input for purposes of the EAL calculations. 2.6.4. Wind Direction UFSAR Section 2.3.2.1.1.1, Figure 2.3-2, Figure 2.3-3, Table 2.3-34 and Table 2.3-130H document the predominant wind direction for GGNS as to the East sector. Thus, a wind direction input of 270° (winds from) is used as the URI inp.ut for purposes of the EAL calculations. 2.6.5. Other Parameters No precipitation is assumed to occur for the duration of the release and plume transport across the Emergency Planning Zone. EP-CALC-GGNS-1701 Page 12 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series
~
- 3. DESIGN INPUTS 3.1. General Constants and Conversion Factors 3.1.1. 472 cc/sec per cfm 3.2. Liquid Effluent 3.2.1. Liquid Effluent Monitor Ranges (UFSAR Table 11.5-1)
- 1) Liquid Radwaste Effluent Line ......................................................... 1E+1 to 1E+6 cpm 3.2.2. Liquid Monitor Response Factor - MRF (GIN-2001/01196)
- 1) Liquid Radwaste Effluent Line ..................................................... 3E+8 cpm per µCi/cc 3.2.3. Liquid Effluent Discharge Source Flow (f}
- 1) Maximum discharge flow (Discharge Permits) ................................................ 100 gpm 3.2.4. Liquid Effluent Dilution Flow (A
- 1) Minimum expected dilution flow (Discharge Permits) ................................... 3,500 gpm 3.2.5. Liquid Effluent Source Term Limit (Eq) 10*CFR*20:, ,. *ODCCM Ljq uii;f LJq1:1(cfl:.in11f . *:':<.Limit** .,
(1,1Ci/ml) .. *.* CµCitml> ., Na-24 5.0E-05 5.0E-04 Mn-54 3:0E-05 3.0E-04 Co-60 3.0E-06 3.0E-05 Zn-65 5.0E-06 5.0E-05 Ag-110m 6.0E-06 6.0E-05 Sb-125 3.0E-05 3.0E-04 Cs-134 9.0E-07 9.0E-06 Cs-137 1.0E-06 1.0E-05 3.3. Gaseous Effluent 3.3.1. Monitor Efficiency Factor - Eff 1
- 1) GE Monitors - Low Ran
- e 08-S-03-22 Section 6.2.2
- Containment Vent
- Fuel Handing Area Vent 7.69E-8 µCi/cc per cpm
- Radwaste Building Vent
- Turbine Building Vent .
EP-CALC-GGNS-1701 Page 13 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series
- 2) SPING Monitors - Channel 5 Low Ran e 08-S-03-22 Section 6.7.5
- S BGT Exhaust A
- Containment Vent
- Fuel Handing Area Vent 3.54E-8 µCi/cc per cpm
- Radwaste Building Vent
- Turbine Building Vent
- 3) SPING Monitors - Channel 7 Mid-Ran e 08-S-03-22 Section 6.7.7
- S BGT Exhaust A
- Containment Vent
- Radwaste Building Vent 1.15E-4 µCi/cc per cpm
- Turbine Building Vent
- Fuel Handing Area Vent
- 4) AXM Monitors - Channel 4 Mid-Ran e 08-S-03-22 Section 6.6.4
- SBGT Exhaust A
- Contain111ent Vent
- Fuel Handing Area Vent 3.04E-6 µCi/cc per cpm
- Radwaste Building Vent
- Turbine Building Vent
- 5) AXM Monitors - Channel 3 Hi
- S BGT Exhaust A
- Containment Vent
- Radwaste Building Vent 1.69E-3 µCi/cc per cpm
- Turbine Building Vent
- Fuel Handing Area Vent
- 6) Canberra Monitors - Normal Range (08-S-03-22)
- SBGT Exhaust B ............................................................... 3.40E-8 µCi/cc per cpm
- 7) Canberra Monitors - High Range <08-S-03-22),
- SBGT Exhaust B ....... '. ....................................................... 9.90E-6 µCi/cc per cpm EP-CALC-GGNS-1701 Page 14 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 3.3.2. Gaseous Effluent Monitor Ranges
- 1) GE Monitors - Low Range (UFSAR Table 11.5-1)
- Containment Vent
- Radwaste Building Vent 1E+1 to 1E+6 cpm I
- Turbine Building Vent 7.69E-7 to 7.69E-2 µCi/cc
- Fuel Handing Area Vent
- 2) SPING Monitors - Channel 5 Low Range (UFSAR Table 11.5-1)
Note - Ranges in cpm were derived from the UFSAR values given in µCi/cc and the monitor efficiency in section 3.3.1.
- SBGT Exhaust A
- Containment Vent 2.8E+O to 1.7E+6 cpm
- Radwaste Building Vent 1E-7 to 6E-2 µCi/cc
- Turbine Building Vent
- Fuel Handing Area Vent
- 3) SPING Monitors - Channel 7 Mid-Range (UFSAR Table 11.5-1)
Note - Ranges in cpm were derived from the UFSAR values given in µCi/cc and the monitor efficiency in section 3.3.1.
- S BGT Exhaust A
- Containment Vent
- 1. 7E+2 to 3.5E+6 cpm
- Radwaste Building Vent 2E-2 to 4E+2 µCi/cc
- Turbine Building Vent
- Fuel Handing Area Vent
- 4) AXM Monitors - Channel 4 Mid-Range (UFSAR Table 11.5-1)
Note - Ranges in cpm were derived from the UFSAR values given in µCi/cc and the monitor efficiency in section 3.3.1.
- SBGT Exhaust A
- Containment Vent 3.3E+1 to 3.3E+6 cpm
- Radwaste Building Vent 1E-4 to 1E+ 1 µCi/cc
- Turbine Building Vent
- Fuel Handing Area Vent EP-CALC-GGNS-1701 Page 15 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series
- 5) AXM Monitors - Channel 3 High Range (UFSAR Table 11.5-1)
Note - Ranges in cpm were derived from the UFSAR values given in µCi/cc and the monitor efficiency in section 3.3.1.
- SBGT Exhaust A
- Containment Vent 5.9E+3 to 5.9E+ 7 cpm
- Radwaste Building Vent 1E+1 to 1E+5 µCi/cc
- Turbine Building Vent
- Fuel Handing Area Vent
- 6) Canberra Monitors - Normal Range (EC 57863 pages 5148-5162)
- SBGT Exhaust B ....................................................................... 1E+O to 1E+9 cpm
- 7) Canberra Monitors - High Range (EC 57863 pages 5148-5162)
- SBGT Exhaust B ....................................................................... 1E+O to 5E+9 cpm 3.3.3. AU1 .1 Gaseous Effluent Source Flow - f
- 1) s*BGT Exhaust A and B (08-S-03-22) .............................. :............................ 4,300 cfm
- 2) Containment Vent (08-S-03-22) .................................................................... 6,000 cfm
- 3) Fuel Handing Area Vent (08-S-03-22) ......................................................... 31,000 cfm
- 4) Radwaste Building Vent (08-S-03-22) ......................................... :............... 48,000 cfm
- 5) Turbine Building Vent (08-S-03-22) ............................................................. 15,000 cfm 3.3.4. AA1 .1, AS1 .1 and AG1 .1 Gaseous Effluent Source Flow - f
- 1) SBGT Exhaust A and B (SFD1102) .............................................................. 4,000 cfm
- 2) Containment Vent (SFD1100) ....................................................................... 6,000 cfm
- 3) Fuel Handing Area Vent (SFD1104A) ......................................................... 24,720 cfm
- 4) Radwaste Building Vent (SFD004 7) ............................................................ 52,495 cfm
- 5) Turbine Building Vent (SFD1105A) ............................................................... 5,000 cfm 3.3.5. AU1 .1 X/Q Dispersion Factor (ODCM Table 2.2.3a)
- 1) All Ground Level Release Points .............................................................4.1 E-6 sec/m 3 EP-CALC-GGNS-1701 Page 16 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 3.3.6. ODCM Dose Factors (ODCM Table 2.1-1) V "."'" Body Ki 13-Skin Li *V-:-AirMi (mrem/yr per µCi/m 3 ) (mrem/yr per µCi/m 3 ) (mrad/yr per µCi/m 3 ) Ar-41 8.84E+03 2.69E+03 9.30E+03 Kr-83m 7.56E-02 O.OOE+OO 1.93E+01 Kr-85m 1.17E+03 1.46E+03 1.23E+03 Kr-85 1.61 E+01 1.34E+03 1.72E+01 Kr-87 5.92E+03 9.73E+03 6.17E+03 Kr-88 1.47E+04 2.37E+03 1.52E+04 Kr-89 1.66E+04 1.01 E+04 1.73E+04
,: Kr-90 1.56E+04 7.29E+03 1.63E+04 Xe-131m 9.15E+01 4.76E+02 1.56E+02 Xe-133m 2.51 E+02 9.94E+02 3.27E+02 Xe-1.33 2.94E+02 3.06E+02 3.53E+02 Xe-135m 3.12E+03 7.11E+02 3.36E+03 Xe.;;135* 1.81 E+03 1.86E+03 1.92E+03 Xe-137 1.42E+03 1.22E+04 1.51 E+03 Xe-.138 8.83E+03 4.13E+03 9.21 E+03 EP-CALC-GGNS-1701 Page 17 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series
- 4. CALCULATIONS 4.1. AU1 .1 Liquid Release 4.1.1. Liquid Effluent Monitor ODCM Limit (derived from ODCM Section 1.1.1)
Per ODCM 1.1.1. Step 2, SF is a normally applied administrative safety factor which causes the calculated Dilution Factor to be two (2) times larger than the dilution factor required for compliance with 1Ox 10CFR20 limits. The normalized activity mix multiplied by their 1Ox 10CFR20 limit by definition yield a sum concentration activity equivalent to the 1Ox 10CFR20 limit, or a compliance dilution factor of 1. compliance dilution factor=[~ Jt0
]
SF=[~:tu}x2 Where: SF administrative safety factor Cg normalized gamma emitter effluent concentration value (fraction) ECg ODCM limit (µCi/ml) Where: SP
- radiation monitor setpoint equivalent to the ODCM limit ( cpm)
Cg normalized gamma emitter effluent concentration value (fraction) ECg ODCM limit (µCi/ml) F minimum setpoint dilution flow (gpm) =0.9 x minimum expected dilution flow SF administrative safety factor - see above f maximum setpoint discharge flow (gpm) =0.9 x maximum discharge flow MRF Monitor Response Factor (cpm/µCi/ml) 4.1.2. AU1 ~ 1 Liquid Release EAL Threshold AU1 .1 liquid is two times (2x) the calculated ODCM limit setpoint. See Attachment 1 for the spreadsheet calculations that develop the AU1 .1 liquid effluent 1 EAL threshold values for each applicable monitor. EP-CALC-GGNS-1701 Page 18 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 4.2. AU1 .1 Gaseous Release 4.2.1. Gaseous Release at the ODCM Limit Where: SP radiation monitor setpoint equivalent to the ODCM limit (cpm) 500/3000 Dose Limit - 500 whole body or 3000 skin (mrem/yr) 472 conversion factor (cc/ft 3 per sec/min) f vent flow (cfm) XIQRP highest land annual average dispersion factor for the release point (sec/m 3 ) Qi activity released (fraction - unit less) Ki whole body dose correction factor (mrem/yr per µCi/m 3 ) Li + 1.1 Mi skin dose factor (mrem/yr per µCi/m 3 ) Eff detector efficiency (µCi/cc per cpm) 4.2.2. AU1 .1 Gaseous Release EAL Threshold AU 1.1 gaseous is two times (2x) the lesser of the calculated whole body or skin ODCM limit setpoint. See Attachment 2 for the spreadsheet calculations that develop the AU1 .1 gaseous effluent EAL threshold values for each applicable monitor. 4.3. AA1.1. AS1.1 and AG1.1 Gaseous Release 4.3.1. Canberra Monitors The AA 1.1, AS 1.1 and AG 1.1 gaseous release EAL thresholds for the Canberra monitors are derived from the SBGT A results as follows: RR (Cifsec) X 1£6 (µCifCi) . SBGT B ( cpm ) =- -------------------- Flow (cfm) x 472 (cc/ sec per cfm) x Eff(µCifcc per cpm) Refer to Attachment 3 for the results of the SBGT B gaseous effluent EAL threshold calculations. EP-CALC-GGNS-1701 Page 19 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 4.3.2. SPING and AXM Monitors The AA 1.1, AS1 .1 and AG1 .1 gaseous release EAL thresholds for the SPING and AXM monitors are developed using the site specific URI .dose assessment model with the inputs described in Section 2 above. Refer to Attachment 4 for the results of the URI gaseous effluent EAL threshold
. calculations.
\
EP-CALC-GGNS-1701 Page 20 of 71 Revision O
GGNS 'EAL Technical Bases Calculations - Ax1 Effluent Series
- 5. CONCLUSIONS 5.1. Effluent Monitor Reading Results in CPM (All Calculated Values Within Range)
.GE . . SAE**- . Alert * . -UE . *.:
Release Point\ Monitor** (cpm) . . .(cpni), .. (cpm) .. *. (c'pm.) SPING 5 NIA .....................................................................................................................................
..................................................................................................................................... NIA NIA 9.32E+5 SPING 7 3.73E+6 3.73E+5 3.74E+4 2.87E+2
*.. *11SBGT A ,
*1 AXM 4 -............................-.-*..**-*-*....- ...........-......................................................................
-*****-*-*........................... NIA 1.40E+7 1.42E+6 _..... ................
-........................... . ...-......................................... 1.08E+4 AXM 3 2.54E+5 2.54E+4 2.55E+3 NIA Canberra normal NIA NIA 1.17E+8 9. 70E+5
. jSBGT 8 ,
Canberra high 4.03E+7 4.03E+6 4.03E+5 3.33E+3 General Electric NIA NIA NIA 3.07E+5
,, S_P_I_N_G . .s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . NIA..................................NIA..................................NIA........................6.68E+5 _______
(, .; .lContainment Vent _S_PI_NG _ .7.. . . . . . . . . . . . . . . . . . . . . . . . . J_._98_E+6............J . 98E+5. . . . . . . ...1: 98_E+4 _..............2.05E+2........
- , *' AXM 4 r*'.* . , ., . . :;I AXM . .i . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.48E+7 7.48E+6 7.48E+5
. . . . 1-*:*35i~+*s*. * * * *********1"'j"5'E"+"4******* . . . . .1. .~3'5""E+*i***** . . . . . . . . . NiA****************** 7.77E+3
,*s*j Radwaste 8 u1ld. Ing
'.tn 1 General Electric
. "s'pfr~'cfs"""'"'"'~:-*. * * -* * * *-* * *-* * * * * . . . . . . . . . .N-/A*************-* . . . . . . . . . .NIA NIA NIA...............* .. . . . . . . . .NIA NIA................ . ......3.84E+4 a*~*34*E*+*~i--.. . .
. a> l Ctn.
L~ivent
.r*.:,i'
- ...'.;J:i Turbine 8 uild ing
=-~----\_ _ -~:~~~:;- -{~~~:~- -{~~~:~- ~fL SPING 7 1.72E+4 1.72E+3
.Q.~.r.:!. ~t?.!..J~J.~~!Et~. . . -.. . . . . . . . . . . . . . ~.!.t\............... .. . . . . . . . .~!.!.\............... .. . . . . . . . .~!.!.\................ .. .J..:?.I~.~.?.. . . ..
SPING 5 1.72E+2 NIA ............................................................................................................
....................................................................................................................... NIA NIA NIA _......................................................
2.67E+5 .
;t * \i/1
- .\*** -*i. Vent SPING 7 4.26E+4 4.26E+3 4.26E+2
*--***-*-********-********-********-*........................................... ................................................ ..............................-................... ..........................................._..... ......*-*****--********-********-*** NIA . ***-****.
1' '. : .*,'.; MM. .1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1..&1J~!!t.............1..:.§.JJ~!.?.. . . . . . .J..:&1J~!~:L. . . . . .~.JJ_~-~-~--.. . .
,*:: AXM 3 2.90E+3 2.90E+2 2.90E+1 NIA
~~~~~~---~~~~~~--~~~---~~~--~~~--~~~----11 General
*-*-*-*- Electric NIA ..-*-..-*-* ............*-*******-*-*****---
........-............................ -...........-...**-*-*****-*-*-.. *--..-*..-******--*-*****-*-*- NIA..***-*-*..- ....*-******--*****NIA 5.95E+4
**. *.. .Fuel Hand Ii ng SPING
*--*-*-* 5 . . ***--*. *-*-*. ****-****-*-****- ***-******-*
NIA NIA -......... .................................................. NIA 1.29E+5 -.. ..
,*: (Aux Bldg) Vent .§.P..!. ~.G.. L. . . . . . . . . . . . . . . . . . . . . . . . Jt1~.~!~. . . . . . . .~.:.1.~_;._::.§. . . . . . _.§._.:1.4. ~-~4....... .. . . . . .}:-l1-A. . . . . . . . .
AXM 4 NIA 2.43E+7 2.44E+6 1.50E+3
;:-** _*.,: J AX°rvf"j'"'"'""""'""""""""'"'""-*********-*******-*. * -* ***-****4ja'E+s*****-* ***-****4*:*3a'E"+"4'"'"""" . . . . 4.~3*9. E+":i. . . . . . . . . . . .N.IA*-****-**-***-*
Liquid Effluent NIA NIA NIA 7.33E+5 Monitor EP-CALC-GGNS-1701 Page 21 of 71 Revision o
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 5.2. Effluent Monitor Reading Results in CPM (For Applicable EAL Thresholds)
- \:***, .:'.;*(';,~.\, ,';" '.* :, ***** ::.* ,.
- ,. :>ReteasePointt::
*:.*,*.':,.St::*>'., , .* f. .. ,; ....
SPING 5 t*;; .:',ISBGTA SPING
*-*****-*-* 7-...........-........................-*-**-*-* .........................- ..................-,..- .......................
.......-.................. 3.73E+5 3.74E+4
--........-.................. ................ 2.87E+2 .
,; . i AXM 3 f; ,:;*?ii~- - - - - - - + - - - - - - , - - - - - + - - - - - - + - - - - - - - 1 1 - - - - - + - - - - - - - - 1 1 2.54E+5
'. *r. :i:j s BG T B .9.§.D..9..~If..§...D.9..fED..§!L____. ___....._. . . . -.. . . ._____.,_____. ____. . ._. ..-.. . . ._. . . . ___________. ____________ . . . -.. . . .-.. . . .-.. . . .________________ . . . . .~.:?.Q_~-~.?.._. . .
* -' Canberra hiQh 4.03E+7 4.03E+6 4.03E+5
':.*.* * ,.* ,j S PI_NG . _5_. . . .*-*--*---*-..-*-*-*-*-*-*-..-* ___. ________. ______________. _________. . . . -.. . . ,____. ______________ . . . . . . . . -.. . . .-.. -.. -..*---*-----*-- . . . . . 6. 68E+ 5______
;i'.*,***f/J.*f
> :.jContainment Vent *-----*-*-, SPING 7-........_,____..____...._,__....___________ ...-........-........____________...._,_______ ...-........1.98E+5
.......-........-........ -...,....___........-....*-*-..---........-....1.98E+4 f'.?grJ AXM 3 1.35E+5
- Cl>'* i* Radwaste Bu .IId . Ing __SPING 5 . ____________. ___________. _____. _____ . .-.. . . .-.. . . .-.. . . .___________ . .-.. . . -.. . . ________________
. __________. ._. ________________ . . ___ . . . -.. . . .-.. . . .-.. . . .-.. . . .________ . . . ______8.34E+4 . ___. . . . -.. .-..---*-**---..
0
- . *~ **i Vent .§.P.!.~.G_L. . . ._________. . ___. ________________. _ _ _ _ _ _ _. . _ _ _ _ _ . . . . _____. 1:.?_?._~::~-----* . . . J . :.?~-~::-~------* . . .___. _ _ _ . . . . -.. . . .-.. . -.. . . .
... AXM 3 1.17E+3
~'ilTurbine Building
;"::,:. :"*:>i Vent
~{~%f------------- _4.26E+3___ --4~2682- ....?Ji?E':~-
k/*;_;....;
.SJ (Aux f:;r<:,_X,X
--------f-_. P-.~-._ _G-~_-.
Fuel Handling BId g)* Vent .SPING
. . . . ._. . . . _. . . . -.. 7
§-_. . .-..___-_____-._. . -. ___.-.. . _- _.-._. . -~.:-.~--~-E-
. . _... . --~-..~-.--t_...-...--...._.-.......---------------------...."""'.__ -......---.......-.-....-.......-.......--......-..____-;.... 1-......-J-.:?-..~---~--~_-§__. . . ... .
. . .-.. . . .-.. . . ._____. ,_,________. . ___ . .-.. . . .-.. . . ._,_. ________. _____. . _ . .-.. .6.43E+5
. .____________. . ______. ___. _____ . . . -..6.44E+4 . . .-.. . . ._. . . . -.. . . .-........ . . . ___. . . . . . . . . -.. . . .-.. . .-......
u'.*,.,,,, AXM 3 4.38E+5 Liquid Effluent N/A N/A N/A 7.33E+5 Monitor 5.3. Gaseous Effluent Monitor Reading Results in Ci/sec
.'JSBGT A
\ <JssGT B 8.1E+2 8.1E+1 8.1E+O 6.7E-2
-';....- - - - - - - + - - - - - - - 1 1 - - - - - - t - - - - - t - - - - - - t
~/~JContainment Vent 6.4E+2 6.4E+1 6.4E+O 6.7E-2 r{8*1Radwaste Building 5.1E+1 5.1E+O 5.1E-1 6.7E-2 i\'11Vent
- :~ -;Turbine Building 1.3E-1 6.7E-2 1.3E+1 1.3E+O 1,., *. !Vent
{:,{ Fuel Handling 8.6E+3 8.6E+2 8.6E+1 6.7E-2
'<?i (Aux Bldg) Vent Note UE release rate values assume a common source term and are two times (2x) the lesser of the Attachment 2 calculated total body or skin ODCM limit setpoint in Ci/sec.
EP-CALC-GGNS-1701 Page 22 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent Series 6~ REFERENCES 6.1. NEi 99-01 Revision 6, Methodology for Development of Emergency Action Levels, November 2012 6.2. 10 CFR 20 Appendix B Table 2 Column 2 6.3. EPA-400-R-92-001, Manual of Protective Actions for Nuclear Incidents, May 1992 6.4. NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 6.5. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, February 1995 6.6. NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012 6.7. Grand Gulf Nuclear Station Offsite Dose Calculation Manual, LBDCR 15051, January 2017 6.8. Unified RASCAL Interface Requirement Specification, Draft 051611 6.9. Unified RASCAL Interface Requirement Specification Grand Gulf Site Annex, Version 2, Draft 022414 6.10. Grand Gulf Nuclear Station UFSAR
- 2.3.2.1.1.1, Wind Distributions {All Meteorological Conditions), Revision O
- 9.4.4.2, Turbine Building Ventilation System - System Description, Revision 10
*, 11.2.1.1, Liquid Radwaste System - Power Generation Design Bases, LDC 05074
- Table 2.3-34, Percentage Frequency of Wind Direction and Speed at Grand Gulf Site Period of Record - August 1972 to July 1973, Revision O
- Tables 2.3-130A through 130G, Frequency Distribution For Pasquill Stability Class A-G, Revision O
- Tables 2.3-130H, Frequency Distribution 1972 to 1976, Revision O
- Table 11.2-10, Liquid Effluents Annual Releases to Discharge Canal, LBDCR 13002
- Table 11.3-9, Expected Annual Release of Gaseous Effluents, LBDCR 13002
- Table 11.5-1, Process and Effluent Radioactivity Monitoring Systems Revision 2016-00 0
- Figure 2.3-2, Annual Wind Rose August 1972 - July 1973, Revision O
- Figure 2.3-3, Comparison of Wind Directions and Speeds at Grand Gulf, Miss., 1972
- 1974 and at Jackson, Miss., 1960 - 1964, Revision O EP-CALC-GGNS-1701 Page 23 of 71 Revision O
GGNS EAL Technical Bases Calculations - Ax1 Effluent $eries 6.11. GIN-2001/01196, Liquid Process and Liquid Effluent Radiation Monitor Calibration Basis 6.12. GIN-2007 /00076, Review of 2001-2005 Annual Average Relative Concentration and Relative Deposition . 6.13. 08:..8-03-22, Installed Radiation Monitoring System Alarm Setpoint Determination and Control, TCN O12 6.14. 06-HP-1 Dl7-V-0001, Turbine Building Occasional release point Monitoring Instrumentation, Revision 101 6.15. EC 57863, Replacement of SBGT "B" Radiation Monitoring System, Revision O 6.16. SFD1102, System Flow Diagram - Standby Gas Treatment System, Revision 003 6.17. SFD1100, System Flow Diagram - Containment Cooling System, Revision 006 6.18. SFD1104A, System Flow Diagram - Fuel Handling Area Ventilation System, Revision 006 . 6.19. SFD0047, System Flow Diagram - Radwaste Building Ventilation System, Sheet 3, Revision 008 6.20. SFD1105A, Turbine Building Ventil_ation System, Revision 012 6.21. Liquid Release Permit Report; Permit Number 2017007 6.22. Liquid Release Permit Report, Permit Number 2017008 6.23. Liquid Release Permit Report, Permit Number 2017009 EP-CALC-GGNS-1701 Page 24 of 71 Revision O AU1 .1 Liquid Effluent EAL Calculations r,..' CIO Cl) gJ .
*e......
0 0 Q, Q. ..... 0 0 0 Q) 0) 0 0
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0 T"' T"' 0 """ T"' 0 CIS
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. Q)
N N* i..: 0' Cl) sCk: '"C
*s E *-
*- ..J
..JO
- 2E
(.)
.8 ~ i5
.c E
- s E
- s E.:.
- s "C
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*-g._
c ' 0 tT
- i .
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- i (.) ::::
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"'- c-
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i~ E:::: e*
~.
- t::
- e::::
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. :t:: E
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(.) Q) ::::1. -
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- 1.
- 0 CIS *-
Eu s...* CIS 0 .... Ck: -
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.. CIS (.)
.... ::L. . .
a.-* (I)
- a. - a. - .I- - z~ 0 T"'.-
- 1. 0 ;!:. w LL*-
Na-24 8.75E-08 8.75E-08 1.38E-02 5.00E-05 5.00E-04 6.88E-06 Mn-54 9.53E-08 3.21 E-08 3.35E-07 4.62E-07 7.27E-02 3.00E-05 3.00E-04 2.18E-05 Co:-60 6.17E-07 4.23E-07 1.27E-06 2.31 E-06 3.63E-01 3.00E-06 3.00E-05 1.09E-05 Zn.;o5 3.04E-07 4.01 E-07 1.32E-06 2.03E-06 3.18E-01 5.00E-06 5.00E-05 1.59E-05 Ag-110m 2.51 E-07 4.58E-08 2.97E-07 4.66E-02 6.00E-06 6.00E-05 2.BOE-06 Sb-125 1.71 E-07 1.71 E-07 2.69E-02 3.00E-05 3.00E-04 8.06E-06 Cs-134 3.92E-07 3.92E-07 6.16E-02 9.00E-07 9.00E-06 5.54E-07 .
.. Cs-1.37 6.18E-07 6.18E-07 9.71E-02 1.00E-06 1.00E-05 9.71E-07 1.27E-06 9.89E-07 4.11 E-06 6.36E-06 1.00E+OO 6.79E-05
-~.*
c
*-***u.
ti ;: e*:u..2
- s E *-
*s
,c ,.= t: -E.CL
.~c~*
Li uid Radwaste Monitor 9.00E+01 3.15E+03 Maximum Discharge Flow (gpm): 1.0.0E+Oi Minimum Dilution Flow (gpm): . 3.50E+03 Monitor Response Factor - MRF (cprn/µCi/mQ: Safety Factor:.* I 3.ooE:osl
* * ; 2_
EP-CALC-GGNS-1701 Page 25 of 71 Revision O AU1 .1 Gaseous Effluent EAL Calculations
\,:}':A,f:~41' 8.8E+03 2.7E+03 9.3E+03 7.2E+01 9.0E-03 7.93E+01 1.16E+02 J(f.:83m 7.6E-02 O.OE+OO 1.9E+01
,.. *Kr"85111 1.2E+03 1.5E+03 1.2E+03 8.8E+01 1.1E-02 1.28E+01 3.08E+01
*/;:::)Kr:-:8.5 1.6E+01 1.3E+03 1.7E+01 3.9E+02 4.9E-02 7.82E-01 6.60E+01 9.7E+03 6.2E+03 6.3E+01 7.8E-03 ,4.64E+01 1.30E+02 i;:<'J(r!88 1. 5E+04 2.4E+03 1.5E+04 9.8E+01 1.2E-02 1.79E+02 2.33E+02 1.0E+04 1.7E+04 6.1E+02 7.6E-02 1.26E+03 2.21E+03 7.3E+03 1.6E+04 4.86+02 1.'6E+02 2.0E+01 2.5E-03 2.28E-01. 1.61E+OO
,i.X~1:33m 2. 5E+02 9.9E+02 3.3E+02
*-:vxel1*3a 2. 9E+o2 3.1E+02 3.5E+02 2.2E+03 2.7E-01 8.05E+01 1.90E+02
>xE:i21_3sm 3.1 E+o3 7.1E+02 3.4E+03 9.9E+02 1.2E-01 3.85E+02 5.43E+02 1.9E+03 1.9E+03 1.2E+03 1.5E-01 2.70E+02 5.94E+02 1.2E+04 1.5E+03 1.3E+03 1.6E-01 2.30E+02 2.24E+03
;;i'XE!-:138 8.8E+03 4.1E+03 9.2E+03 1.0E+03 1.2E-01 1.10E+03 1.78E+03 8.0E+03 1.0E+OO 3'.64E+03 * *S:-14E+03*.
Calculation Constants Cont FHA RW TB Disperaion-~Q(sedm~:~~~~~~~~~4-~_0_E_:o_6__._~_'.4_._1o__E_~~---t_:*_~~--~-io_t_;_.~_i_:_~_~_1_0E_-t_B6_-_~ Effluent Flow- f (cfm): * *. 6._00E+03 . *3;,10E+'64:.' :,4*:.8c5E-t04.* *-:{50Ef,Q4\. GE Eff (µCi/cc I cpm): ;:l,;SQE-08 * * . 7::~9E/Oa.:*. ;,:. *i: : '::. ,:Z.15$,E:"Cl8;.*, SPING Ch 5 Eff (µCi/cc I cpm): /::3':54E-'Oa:* *_:3.54E~;oe_, \ :'J3.:$4.E~Qtf\ d<t54E:JQff/ ;{:~f$4E::;08'( SPING Ch 7 Eff (µCi/cc I cpm): {:15E-04. ,*. i*15E,i04. '.:1'.15E:.04< ,;;',1~,15Ei~t,:( [, *1:,15~do4:* AXM Ch 4 Eff (µCi/cc I cpm): ;: . E':: 3:'64E:06i: Canberra Norm Eff (µCi/cc I cpm): Canberra High Eff (µCi/cc I cpm):
~~~ ~~~ ~:~: ~~~=:~;~:! : :0:0~1 EP-CALC-GGNS-1701 Page 26 of 71 Revision O AU1 .1 Gaseous Effluent EAL Calculations Calculated Seteoint Results SBGTA SBGTB Cont FHA RW TB SP-TB (µCi/sec): 3.35E+04 . 3.35E+04. 3.3.5Et04 3.35E+04 3.35E+04. 3.35E+04 SP-Skin (µCi/sec): 8.99E+04 8.99E+04 8.99E+04. 8.99E+04 8.99E+04 8.99E+04 SP-TB (µCi/cc): 1.65E-02 . 1.65E-02 1. ~8E-02
- 2.29E-03. 1.48E-03 4.73E-03 SP-Skin (µCi/cc): 4.43E-02 4.43E... Q2 3.18E-02 6.15E-03 3.97E-03 ,1.27E-02 GE SP-IB (cpm): 1.54E+Q5. 2.97E+04 1.92E+04 6.15E+04 GE SP-Skin cpm): T99E+d4 5 ..16E+04. 1.65E+05 SPING Ch 5 SP-TB (cpm):
- 6 . 46E+04 4.17E+04 1.34E+05 SPING Ch 5 SP-Skin (cpm): 1.74E+05 1.12E+05 3.59E+05 SPING Ch 7 SP-TB (cpm): 1.99E+01 1.2aE+01 4. 1'1E+01 SPING Ch 7 SP-Skin c m :
- f5:35E+01: 3i45E+01 1.10E+02 AXM Ch 4 SP-TB (cpm): ___ .....--................ ----.~--* 7.52E+02
- 4.86E+b2 . 1.55E+03 AXM Ch 4 SP-Skin c m : IIDmB IDml&!IIIIFmm!Dllllmll.Ulllmlml
- : _______ lllmilB
- ~
Calculated UE Threshold Results
*.GE AU1.1 (cpm):
SPING Ch 5 AU1.1 (cpm): 1--....-;.-......_..,.. SPING Ch 7 AU1.1 (cpm): -_.;:;....;.,...-....;.. AXM Ch 4 AU1.1 (cpm):
~anberra Norm AU1. 1 ( cpm):
Canberra Hi h AU1.1 c m: EP-CALC-GGNS-1701 Page 27 of 71 Revision O
\,
SBGT B Canberra Monitor Correlation GE SAE Alert (AG1.1) (AS1.1) (At\1.1) Canberra Norm (cpm): >c1.17E+fo.; * ;1.17E~:0.9{) '.,1ifi~1lEt'.OSi1 Canberra High (cpm): \4:\0$EfO:i~t tf4:03E¥Q'(i;1!t ~,~t'031;+:Q5~~ SBGT SPING 7 Release Rate@ General Emergency (Ci/sec): *:B:09E'.f02, UCF (cc/sec per cfm): ,,.;,.,, A7'2'/;;::t UCF (µCi per Ci): i'fOOEfO~?\ Effluent Flow- f (cfm): AJ3QE+.03> Normal Channel Eff (µCi/cc/cpm): ,'3AOE~Qff, High Channel Eff (µCi/cc/cpm): , 9.90_6106>'. EP-CALC-GGNS-1701 Page 28 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT SPING 7 - General Emergency Dose Assessment Grnnd Gulf Tuesday, February 6, 2018 20:30 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS><Supp><Contain><Aux ffldg> ..:SBGT><Env.=- PRf: *t _60E-05 Containment Hl..JT: = < 2 Hours Contairiment Spr.a,s:. = OFF Supp Poof S1.atus: =Subcooled Safety Fi!t;,rs; =Wor',;ing HVAC Fib;;rs: = NIA A!.:Jx Bldg HUT:="": 2 Hours Turbine Bldg HUT: =N.fA Radl/lfaste Bldg HI.JT: =Ni,.!i. Source Term: Reactor Core Accident - Clad OnSite Lower Time After SID (hh:mm): 1:00 W'ind: From 27D" @ 4.4 mph Relea&e Duration {hh:mm): 1:00 ETE (hh:mm): [NJA] :Stability Ch:,ss: D Precipitatton: None Monitor: SPING ch 7 Readings: 3.73E+06 cpm flm.wate: 400D CFM Distance Exposur~ extern.JI lnh.:.;!ation Deposition T.EDE COE Evacuation Are::'IB From O:tr) 10 Miles Rate *Plume.ODE CEDE Ground ODE Thyroid Circle distances are 2 . .5 and ,O mil.es.. (Mtlesi *tmR/hr) *,(rnRem) (mRem} (inReni) (n~R~di) (mRem) A
.., * *'s:s_. '\1aOOE+03 .*. 6_92E+01 R B 1.43E+D3 9,99E.+.02 2.73E+OO *t.QSE+OO 0.5 1."i9E+03 8.32E+02 1.98E+OO 7.71E-D1 .B.35E+02 5.04E+0*1 0.7 8.44E+D2 5.84E+02 1:16E+01 9.49E+OD £.O=*E+02 3_ 15E+0*1 "1.0 5.56E+D2 3.81E+02 1.12E+01 9.29E+OO 4-il2E+02 *t.83E+01 1.5 3.29E+02 2.24E+D2 9.40E+OO 7-54E+OO 2.41E+02 9..88E+OD p D 2.0 2.59E+02 1_76E+D2 7.52E+OO S-77E+OD *L9UE+02 6.?4E+OO 3_0 1.82E+02 *1-24E+02 3.87E+OO 2.94E+OD 1.31E+02 3.60E+OD 4.0 1.51E+D2 1_03E+02 3_66E+OO 2.69E+OO 1.09E+02 3.03E+OO B.2BE+O'l 2.64E+OO N E 5_0 1.30E+D2 8.70E+D1 3-37E+OO 2.39E+OO 7_0 9'.80E+01 6_67E+D1 2J32E+OO 1.8BE+OO 7_14E+01 2.14E+OO 1fto 6.72E+D1 4.52E+0*1 *L99E+OO *t.23E+OO 4.8.;E+O*l *t_59E+OO A!:,sessment Data Resul1S S.a.ved to :-ile: M F Grand Gulf 10Miles Mcn~orerl Rel-e~se 02G6W18 .203D34.URl7 K H J
Particukii.e i .2SE-03 (0.0%} Iodine 2.4 lE-02 (0,0%) Reviewed By: Gr;;JKII Gulf l 2:.0 , .0 EP-CALC-GGNS-1701 Page 29 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT SPING 7 - Site Area Emergency Dose Assessme:nt Grand Gulf Method: _Detailed Assessment - Monitored Release Release Pattri.vay: <RCS><.Supp><Contain><Aux Bldg><SBGT>,<Env> PRF.: UOE-05 Containme.ntHUT: =< 2- Hours Containment Sprays: = OFF Supp Poot Status: =Suboooled Safet/Filt~rs:= Wc,r&;'.ii:lg _ HVAC*Filters: = fiiA Au:,:Blclg:HUT: = <21-:!ours TurbJnieBldg HUT: =N!A RadWa.ste Bldg HUT: = NIA Source Tem1: Reactor Core Accident - Clad OnSite Lower Time AfterstD-(hh:mm): 1:00 Wind:'From 270"@.AA mph Release Duration (hh:mm): -1 :00 ETE {hh:mn'i j: {NJA ] Stability CEass: D Precipitation: None Monitor:SPING ch 7 Readings, 3-7.3E+OS cpm Ffowrate: 4000 CfM
._:~-,_.:,:Sl_-~---_--7:*:','.J>,.;~*?.:*J=~:c"ij,:,,,n.,ft_,[:?~
fF_;'\.,].'.':-\.:{-Q;,**,--'~l~-~~)~~.r-*0_1}:,<.1~;f;; ,t ~i'~-,~,.~ti *_~.*-.i,~-~,,_'~*-w(. . e~__r~,~1~-J"'-G.,i~,*,.~t:w~l).t~' c.
.. 77 1
1 __ .,u,,2. ~t'.:~,;~'i;p-.:~~~I:J:.-\,'~(-iL:-"--"....* ....,....:_:/_]~_-va....'c_.u_a_'t1,...*c~_*n_:A_l_:e_.~*...s;....-lf....fo_n_)_O_tt_&_'}'1_*0...,;~_,,_-il_es_?_'<__-_::,;_.,--_-___, i;;cc':!-'I.
,,.-:'* **:\*It"_?~.:-,~;,_ _ ":~: , ,. , * -*' - , _ ;_ __ 'A, , .?' \.: ,.~. -,,,,.,..... Circle dfs.tances Sr.(:: 2. 5 and 1.D mlfes.
d ~,*;.:;,: ,:'\'.:"., " ., - :.- -**.,. , , ti-,, ,, , F}-i"' . --- R A
.. --~JL ... , .,J,1.;i~.~Pl.. ___ft,~~~~-AJ ....,.~-Z~J::;:QJ, _J,Q§J;~J ... , ..1_,QJ11:;t9~, .. t?~-~?~tJl:9. B 0.5 1.19E+D2 8.32E+01 1.98E-01 O.OOE+OO 8.34E+01 5.04E+OO 0.7 8.44E+D1 5_84E+01 *u~E+-00 9.49E-01 6-05E+01 3.15E+OO 1.D 5.56E+01 3.81E+D.1 1.12E+-DO '.9.29E'--01 -4.02E+O"l t~838+00 1.5 3-.29_E+01 2.24E+D1 9.40E-01 7.54E-01 2:41E+01 9:8SE-01 p D*
2.0 2.59E+D1 1-76E+D1 7.52E-01 5.77E-£n *1.9-DE+01 . 6.24E-D1 3.0 1.82E+01 1.24E+01 3.87E-01 2:94-E-01 1.31E+01 4.0 1.51E+01 1'.03E+01 3.66E---01 2.59E-01 1.09E+o*t
- S.70E+OO 3.37E---01 9;:2.SE+OO :2.64E~01 N E 5.0 1.30E+01 2.39E-O' 7'.0 9.80E+OO 6,67E+DO 2,82E---01 1,88E---01 7.14E+OO _2:14E~01 10.0 6.72E+DO 4.52E+DO *L99E---01 'i_.23E-0.1 4.84E+.OO 1:ssE.:01 Asses~menfData Resul~ Sa11ed to File; F G.rancl'Gu.lf 1Di1iles l\ilonitored Rel-ease 020--fl21J18 203603JJR17 H
J Reviewed By: EP-CALC-GGNS-1701 Page 30 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT SPING 7 - Alert Dose Assessment Grnnd Gulf Tuesday, February 6, 2018 20:JB Method: DetaHed Assessment - Monitored Release Release Path*~vay: <RC.S><Supp><Co11tai11><Aux B!dg><SBGT><Env.::- PRf: *LGOE-05 Contai.nme,nt HUT: =--::: 2 Hours Con:tainment Spr3:,,.;;: = OFF Sup.p Pool S,atus: = Subcoolo:d Safety Filters:= \Norking HV .&.C Filters: = N,IA A1.;x.Bldg HUT:= <:.2 Hours Tu,bln,e Bldg HUT: =N.!A R;;,dW.aslte B!d,g HIJT: =N/,o. Source Term: Reactor Core Accident - Clad OnSite Lower Time After sm (hh:mrn): 1:OO Wind: from 27u @,4.4 mph 0 Release Duration {hh:mm): 1:DO ETE (llh:mm ): [NIA] :StaiJility Crass: D Precipitat1ori: None Monitor. SPING 1 ch 7 Readings: 3.74E+04 cprn fiowrate: 400D CfM
- Distance Etpoiure ~:.;terh:.;l lnh~lation Deposition *rnp;E CC!E Eva,cuatiN1 Are,as fromO to-1'0 Miles 1
Rate PJun~e,DQE . t:;EDE Groun*d.bDE ' :Thyrofi;f: Circle distances are 2 ..5 and 1;!] miles. (Mflesj (mRlhr) (mR.em.:- . (mR.em} (inRem;, '(nJ;f\eml (n,Rem) ', A S.B. O.DOE.+00 D.OOE+QO 1.0DE+O_t R 8 1.44.EtD1 1 ..00E+01 6,93E~_Q1. 0.5 1.20E+01 8.36E+OIJ 0.00:E+OO O.OOE+OQ EUBE+OO 5.04E~D1 0.7 S.48E+OO 5.88E+OO U6E-f41 O.ODE+Ol} B.ODE+OO 3.16E~D1
'1.0 5.60E+DO 3.84E+OO 1.14E-Dl 0.00E+OO 3.95E+OO 1.84E-01 1.5 3.32E+OO 2-26E+OO O.OOE+OO O.ODE+OD 2.26E+O-O {LOOE+OO p D 2.D 2.61E+OO 1.78E+DO O.DOE+OO O.ODE+OO 1.78.E+OO {toOE+OO 3.0 1.84E+OO *1.2SE+DO O.OOE+OO O.OOE+OO 1.25E+OO O.OOE+OO 4.0 1.52E+OO 1.04E+OO U.OOE+OO O.OOE+OO 1.04E+OO O:ODE+OO 5.0 1.31E+OO 8.76E-01 O.OOE+OQ O.OOE+OO ,9.76E-01 O,OIJE+OO N E 7.0 9.BBE-01 6.72E-01 O.OOE+OO O.OOE+OO 6.72E-01 lJicOOE+OO
*rn.o 6.7£E-01 4.55E-01 O.OOE+OO 0.0.DE+OO 4.55E-01 D.OOE+OO Ass1:-ssment Data Resul:is Sa.vecl to File: M F Grand Gulf 10Miles Monitored R.-e!-ease 02062018 20:?,804.URl7 K H J
Revievved By: t,fo *-le G1.u.1 8, EP-CALC-GGNS-1701 Page 31 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT AXM 3 - General Emera-ency Cose Assessment Grand Gulf Tuesday; February 6. 2018 2lJ:40 Method: DetaHedl Assessment - Monitored Ref ease Release.Pathway: <RCS><Supp><Contain> <Aux B1dg><SBGT><Eo.v> PRf: J _60E~05 Coritainmen! HUT: =< 2 Hours Con.tainment Sprays: =OFF Supp Poot Status~ = Su'be:ooled Safet;{Filti!rs: =Worting HVAC Filters:= NIA A!.Jx Bldg HUT:= <::2 Hours . TumlnE Bltjg KUT: =NlA Radl/ir:astt.e Bldg !-{ITT: =NIA Source Term: Reactor Core Accident - Cl-ad *OnSUe Lower Time After SID (hh:mm): 1:00 Wind: From 27oa @\f.'4'mph Release Duration (hh:mm): 1:DO ETE (hh:mm): IN/A]
- St13bilify .Class'. D Precipitatton= None Monitor.: AXM ch 3 Readings: 2.54E+05 cpm flowrate: 4000 CfM Cir.:.le df-s.tances al"e 2. ,5 and 10 miles.
A R *e
.J,.4}E+O~ "'* .9~~~!;:1;:,Q:;?;. ,,2.J3!:;t00.. ~ 51 + I(. . 6~~:2E+Q1_ ,
05 1.19E+03 8.32Ef02 t'.98E+OO T!E-t: 8_35E+02 5:04E+01 0.7 S.44E+D2 5.B4E+02 t:16E+0,1 9.50E+cm 6-05E+02 3.15E:i.01* 5.56E+02 3.82E+D2 1.13E+01 9.30E+OO **:L0:2E+02 1:83E+01 1.5 3.30E+02 224E+D2 9.40E+OO 7.58E+OO 2At_E+02 9.BBE+OO p D 2.0 2.59E+D2 1]6E+02 7.52E+OO 5-7-BE+OO *L90E+t12 .624E+Cl0 3.0 1.82E+02 1.24E+02 3.87E+OO 2.95E+OO 1-31E+02 4.0 1.51E+D2 t03E+02 3.66E+OO 2_69E+OO 1_09E+02 2.39E+OO !V34E+OO N E 5.0 8.71E+01 3.37E+OO 1.0 9.80E+01 6_67E+01 2.82E+OO 1_88E+OO 7_14E+(U 2.14E+OO 10.0 6.72E+D1 "t99E+OO 1.23E+OD 4J:l4E+01 .t:59E+OO Assessment Da.ta Resulis Saved to File: M F Grand Gulf 10Miles Monitored Ret-ease 020fi21J18 2fl4D45.URl7 K H J Revie...-red By: EP-CALC-GGNS-17 01 Page 32 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT AXM 3 - Site Area Emergency Dose Assessment Grand Gulf Tuesday~ February 6, 2018 20.:42 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS><Supp><Contain><Aux flldg><SBGT><Env> PRf: L60 E-05 Containment HllT: =< 2 Hours Containment Sprays: =OFF Supp P.oo! S~tus: = Su'bc,ooled Sa.fety Filters:= \l'fori;inQ HVAC Filters:= NJA Aux Bldg HUT:=< 2 Hours Turbine Blc!g HUT: =N/A Ra,jW.:;,s;te Bldg HIJT: =NIA Source Term: Reactor Core Accident - Clad OnS1re Lower Time After SID (hh:mmj: 1:00 Wind: from 270°@ 4.4 mph Release Duration (hh:mm): 1:00 ETE (hh:mmJ: [NlA] Stahiliiy Ciass: D Precipitat1.on: None Monitor: AXM ch 3 Readings: 2.54E+04 <::pm flowrate: 4000 CFM Dis:tan:ce Ex~1osur~ _, J:.:~ternal .Inhalation Deposttia:in TEDE CDE Rate Plun_1e,b'D?E.'*_ GEDE GriJrjncfbDE I* Thyroid Circle distancf:'s .:1re 2, !5 and ia miles. _(Mtlesi **(mRthrf (rhRe111if.. (niRernJ UnRem;, JmRem) {m.Rern) A SiJ3, 1.43E+02 .9. 9~E+01 2.73E~01 __ 1.DGEA}1 1.00E+02.. 6.92E+OO R 8 0 0.5 U9E+02 8.32E+01 *].98E-01 O.OOE+OU B.34E+01 5.04E+OO 0] 8.44E+01 5.84E+0*1 U6E+OO 9.SOE-01 B.05E+0*1 3.*15E+OO
*1.0 5.56E+01 3.82E+O*J 1.13E+OIJ 9.30E-D1 4.02E+0*1 1.83E+OO 1.5 3.30E+01 2.24E+0*1 9AOE-01 7.58E-01 2.41E+01 9.88E~D1 p 2.0 2.59E+01 1.76E+D1 7.52E-01 5.78E-iT1 *L90E+O*I f,.24E-01 3.0 *1.82E+D1 1.24E+01 3.87E-01 2.95E-01 1.31E+01 3.60E-0-1 4.0 1.51E+01 1.03E+01 3.£6:EAH 2.69iE-D1 1.09E+O*J 3.03E-0*1 5.0 1.30E+0*1 8.71E+OO 3.37E-flr] 2.39E-01 9.29E+Ofl 2.134E-01 N E 7.0 9.80E+OO 6.67E+OO 2.82E-01 L88E-D1 7.14E+OU 2.14E-01 10.D 6.72E+OO 4.52E+DO 1.99E-01 -1.23E--D1 4.84E+OO 1.59E.. 01 Assessment Data Results Saved to File: M F Grand Guif 10Miles Mcni,ore-0 Rel.ease 02062018 20425i.URJ7 K H J
Porticutate 1.2SE-{M (0.0%) Iodine Reviewed By: Or,?J!;{ll Gui(. I 2.0. t ,0 EP-CALC-GGNS-1701 Page 33 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT AXM 3 -Alert Dose Assessnie'nt Grand Gulf Tuesday; February 6;2018 20:44 Method: DetaHed Assessment - Monitored Refe.ase Release.Pathway: <RCS><Supp><Contain><Aux Bldg><SBGT><Env::- PRf: L60E-05 Coritainment HUT: =< 2 Hours Containment Sprays: =OFF Supp ?ool St,.tus: = Sub,::ooled Safety Filt,ers: =Woridng_ HVAC Filters:= NIA Aux.S.ldg HUT:=< 21-!ours Turbln,e Bldg HUT: =NlA R.;.d!/1/:as~.Bldg HUT: =N,IA Source Term: Reactor Core Accident - Clad OnSite Lower Time After SFD {hl1:mm}: 1:00 Wind: from 270"@JA mph Release Dumtion (hh:nmi}: 1:DO ETE (hh:mm): IN/A J Stability Cii)ss: D PredpitaJt.on: None Monitor: AXM ch ;3 Readings: 2.55E+03 cpm Ffowrate: 400D CfM
- Distance Exposure .* 'J=::~temcit lnh.Jlation : Depo$tticn . *reoE *wt*
-Rate. -8Ju~}eJ::iDE CEDE Gtriun:d 'bDE Th}<l'Qi&y, Circle distances are 2: .5 and 10 mi!,es.
. {~~~hes( * (mRfhr)',: :'(rnRemJ . (n1Rem) . \ihi=t."erf1!; * (nl~r:'1ffi) (n~R.etnt ' A R 8
- S_._R __134.E.~01 .... -.- ..:U1J;~P 1.... .,O._DOE+PO. . ;~;O~J;,-i;()_O_ . ~"q\E":OJ ." 6_.93E:D.1 ...
0.5 1.20E+01 8AOE+OQ O.OOE+OO O.OOE+OO 8.4bE+OO 5.04E-01 0.7 8.48E+OO 5.88.E+OO 1.1GE-01 O.OOE+OO £.ODE+OO 3;16E.,01 1*.o 5 . 60E+DO 3 ..84E+DO 1.14E-{H OJJOE+OO 3.96E+OD 1.84E-01 1.5 3.32E+OO 2.26E+DO O.OOE+OO O.OOE+OO 2.wE+oo* trnOE+OO p D* 2.0 2.61E+OO 1:78E+OO {toOE+OO {lWOE+OO *U8E+OO O,OOE:st:00 3:0 1.84E+OO 1.25E+DO O.OOE+OO O.OOE+OO 1.25E+OO O.OOE:+00. 4.o* 1.52E+OO 1.D4E+OO O.OOE+OO O.OOE+OO 1.04E+OO O:ODE+OO O:.OOE+OO N E 5.0 L31E+OO 8~77E-01 D.OOE+OO O.OOE+OO 8.77E.-01 7.0 9.BSE-01 6.72E-!l1 O.OOE+OO D,OOE+OO 6.72f.-0*1 lhOOE:+-00 10:0 6.7£E-01 4s56E--01 O.OOE+OO D.ODE+OO 4.56E-D*J D,OOE+OO AssessmentData Resuhs Saved to File: M F G*rand Gu1f 10Miles Monitored Release lJ21lf!21l18 20444i.UR17 lcidlne
- 2.42E~04 (0.0%)
Reviewed By: Page 1 of S EP-CALC-GGNS-1701 Page 34 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT AXM 4 - Site Area Emergency Cose Assessment Grand Gulf Tuesday, Februar116, 2018 20:47 Method: Detailed Assessment - Monitored Release Release Pathway: <RCS><Supp><Contain><Aux Bldg><SBGT><Em*~"' PRF: *1 _GOE-05 Containment HUT:=< 2 Hours Containment Spr.:.ys: =OFF Sl.llpp Pool St.;;tus: = Subcooled Safety Filters: =lNorking HVAC fib:rs: = NlA Aux Bldg :HUT:=< 2 HDurs Turblne Bldg HUT: =N/A Ract1il!.as:te Bldg HI.JT: = NIA Source Term: Reactor Core Accident - Clad OnSlte Lower Time After sm (hh:mm): 1:00 1/i/ind: from 27D 0 @ 4.4 mph Release Duration {hh:rnm): 1 :DO ETE (llh::mm j: I~UA ] Stai)ilitiJ Cfa3e;s: D Prei::ipitation: None Monitor: AXM ch 4 Readings: 1.4DE+07 cpm F!owrate: 4000 Cflw1 Dir,t::rnce Exposure, External -** -* lnhcil,:.:ition :Oepo6.ition 'TED,E CDE. Evacuation Are.as from o t,:. 10. Miles Rate Plume_DDE: CEDE Ground'DDE . Thyroid Cim!e distances are 2 *.5 and rn mi!,es. (Mtles) '(mR:/11.r) .(l:'nRem} (mRem) *(mRem) tmR-erri'.i A S.fk 9.98E+0.1 __ 1,00E+02 R B 0.5 1.18E+02 8_32E+01 ROOE+OD 8.34E+01 5_01J;Ei*OO 0.7 8AOE+0*1 5,80E+0*1 1_*16E+OO 9.46E-fl1 6-01E+O*J 3 .. 13E+OO 1.0 5.52E+01 ::rnaE+o1 1.12E+OO 9.26E-0*1 4.01E+O*t 1_82E+OO 1-5 3.28E+01 2-24E+01 9AO!E-01 7 .54E-01 2-41E+01 9.SOE-01 p D 2.0 2.58E+01 1 _76E+01 7.52E-{n 5.77E-01 *LB&E+O*t 6.20E-D1 3_0 1.82E+01 *1_24E+0.1 3.87E-01 2.94:E-01 1_30E+01 3.57E-O*t 4_0 1.50E+01 1_03E+01 3J36E-D1 2.58E-lJ*1 1_09E+Q-1 3.01 E-O*l 5_0 1.30E+o-1 8.68E+OO 2.38E-01 9-25E+OO 2.62E-01 N E 7_0 9.76E+OO 6_66E+OO 2.81E-01 1.87E-D1 7_13E+OO 2.13E-01 1-rrn 6.72E+DO 4S1E+OO *i .23E-D1 4-El3E+OO 1.58E-01 Assessment Data Resull.S Saved to File-: M F G.rand Gulf 10Miles l\,km,rtored R.el.ease 02062018 204719.URl7 K H J Reviewed By: t,Jobte Go,s Or;i.11,:11 Gulf I 2:.0. l .O EP-CALC-GGNS-1701 Page 35 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations SBGT AXM 4 -Alert Dose Assessmeht Grand Gulf Method: Detaifed Assessment - Monitored Release Release.Pathway: <RCS><:Supp><Contain>.::Aux Bldg><SBGT><Env> PRf: *t ~60E-05 Coniainme:rit HUT:=<:* 2 Hours Containment Sprays:= OFF Supp ?ool St3tus, = Suboooled.
- SafefyJFilters:. = 'li.icili:i~g-HVAC Flhers: =NJA Aux Bldg !HUT:= <2 Hours Turbln,eBl¢g HljT;i::-N.fA RaalW.aste .Bldig HUT: =NiA Source Term: Reactor Core Accident - Clad OnSite La\*ter TimeAft:erSID (hh:mm}: 1:CJO Wind: from 27.0° @,4A mph
.Release Duration {hh:mm):-1:00 ETE {hh:mm): [NJA] Stal) ilify Class: D Predpitati_on: None Monitor: AXM ch 4 Reading_s: 1A2E+06 cpm Ffowrate: 400!)-CfltJ1 -
Circle dfstances are 2; .5 and_ t O m if;es..
-A R 8 0.5 1.20E+01 8AQE+OQ O~DQE+OO O.OOE+OO 8.4GE+OO- 5.04E.;01 OJlDE+OO £,OOE+OO 1-o 5.60E+O(I 3._85E+DO 1 ..l4E~01 OJJOE+OO ~.96E+OD*
LS 3.32E+OO 2.26E+DO O;OOE+Dff OJJOE+OO 2.2:SE+OO. O"OOE+OO 2.0 2.61E+OCJ 1]8.E+DO O.DClE+DO {UlOE+OO 't.7~.E+OO t:tOOE:;'.OO 3;0 1.84E+OO 1:25E+OO {toOE+QO o.'r.:HJE+OO 12SJ:+OO O,QOE+OO-4.0 1.52.E+DO L04E+OO O.DOE+DO O.OOE+OO 1.04E+OO O:OIJE+OO OJJOE+OO 8.78E~01 *O:OlJE+OO _E 5.0 1.32E+OO 8.78E-01 O.OOE+DQ 7.0 9.88E-01 6.73E-01 O.OOE+OQ O.ClOE+OO 6.73E:-0*1 O:OOE+OO 10.0 6.76E-01 4.-56E:.0:1 O.OQE+OO O.OUE+OO 4.56E~01 O:OOE+OO
*Assessment Data Results Saved to File: ... F Grand Gulf 1OMiles Monitored .Release 020021118 204944.URl7 K H J
Reviewed. By: Gr~ G1,1lf / z.o. ~;o EP-CALC-GGNS-1701 Page 36 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent SPING 7- General Emergency Dose Assessment Grand Garf Tuesday, Febrmny 6, 2018 20:55 Method: Detaifed Assessment - Monitored Release Release Patl:rway: <RCS><Supp><Cont'.lin> <HVAC fflters:~ <E1w> PRE S.OOE-04 Contai.nme.m HUT:= .:;*2 Ho1;:1rs Gcn:tainment Sprays:= OFF Supp Pool S!L:itus: = Subcooled Safety Filtrs: = NJA HVAC. Filters: = VVorking Aux Bldg HUT: =NiA Turbine Bldg HUT:= WA Rad1Nsste Bldg.HI.ff:= NIA Source Term: Reactor Core, Acddent - Clad OnSite Lr:Vl'ter Time After sm (ht1:mm): 1.:00 Wind: from 270° @.4.4 mph Releas.e Duration {hh:mm): 1 :DO ET:E (il h:m mt [N/A ] Stahilify C!i3$&: D Predpitatton: None Monitor.: SPING ch 7 Readings.: 1.98E+06 cpm *flowrate: 6000 CFM Dist:mce Exposure E:~te:rncJ :
- lnt-1.:Jl,ation Oepos.ttion . ::EDiE CC>l= Evricua:tec,n Are.['ls from, O_to 10. Miles Rate Plume . DD£ CEDE Grcrund.'bDE Thi*roid Circle di;;.tances are 2..5 and tD miles.
(Mfles) (mRfhr) ,(mR,en11J ' t;r'nRemj (inRem) *(mRem) (11,Ren1) A
; R 8
;>S:B~";;.;:,: 1.2'1E+03 B.39E+.02 *1 0 15E+02 .. _4.89!::+0*1 2..TlE+03.,.
0-5 9.92E+02 6J34E+02 8.80E+01 4.04E+O*t 8.12E+02 '1.90E+03 0-7 6.76E+02 4.64E+02 5.84E+01 2:.91E+0*1 5-5:lE+02 L14E+03 1.0 4.28E+D2 2.92E+02 3-71E+D1 1.96E+01 3.49E+02 6.2BE+02 1.5 2.39E+02 *J.62E+02 2.13E+01 1.14E+O'I *J.94E+02 3.2BE+02 p D 2.0 1.88E+02 1.28E+02 1.41E+0*1 7.38E+OO 1.49E+02 2.09E+02 3.0 1.53E+D2 1~05E+02 8.93E+OO 4.58E+OO '1.19E+02 1.4DE+02 4.0 1.24E+02 3.51E+01 7.71E+OO 3.89E+OO 9.67E+01 1.16E+01 7.73E+O*t 9.55E+01 N E 5.0 1.02E+D2 6J7E+0'1 6.42E+OO 3.17E+OO 7.0 7.84E+D1 5.18E+01 526E+OO 2.47E+OD 5.96E+O*t 7.67E+01 lO.O 5.04E+D*1 3.35E+D*1 3.69E+OO 1..61E+OO 3.BBE+O*t 5.75E+0*1 Assessment Data Resulls Sa.11ecl to File: M F Grand Gulf 10Miles Manitore<l Roel.ease 021}62018 205509.URI? K H J Portieulilte 4.98f-*02 (0.0%) Noble G.:m Revie\i\.ied B y : - - - - - - - - - - - - - - - - - - - EP-CALC-GGNS-1701 Page 37 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent SPING 7- Site Area Emergency Dose Assessment Grand Gaff Tu,e,sday; February 6, 2018 21':l::56 Metllod::OetaHed Assessment- Monitored Refease
- Release Pathway: <RCS><Supp>c:Contain> <HVAC Filters> <Env> PRF: KOOE-04 Containmecnt HUT,=< 2* Hours Containment *sprays: =OFF Supp Peto! Sitaitu~: =Sube:ooled Safety'.Filters: = NJA
*HI.IAC Filters: =Working Aux Bldg HUT: =W'.A Turbine Bldg HUT: =NlA RadWas.~ .Bldg r{UT: =Ni A
.Source Tem1: Reactor Core Acddent - Clad OnSite Lower Tinie After SID{hh:mm ): 1:00 . Wind: Frnm 270°@/J.4 mph Release Duration {1'1hch:1m):*1 :00 ETE f.hh:mm): [NIA] - Statimfy Class: D Precipita_![.on: None Monitor.: SPlNG ch 7 Readings: 1.98E+05 cpm Ftowrate: 6000 CFM 0.5 9.92E+01 6 ..84E+01 8.80E+OO 4.04E+OO 8.12E+01 1 anr::: 1 }2 0 .7 6.76E+O 1 4.64E+01 5.84E+OO 2.91E+OO 5:51Ei-01 't.14E+02 1.0 4.28E+01 2 ..92.E+O'l 3.71E+OO 1.96E+OO 3-49E-1:0*J 6.28E+01 1 .5 2.39E+D 1 1.62E+D1 2 .. 13E+DO 1.14E+OO 1.94E+0.1 3.28E+01 p 0 2.0 . 1.88E+01 128E+D1 1A1E+DO 7,38E-01 1.49E+01 2.09E+01 3.0 1.53E+01 1~05E+01 .8.93E-01 4;5iE-01 1JSE+0'1 1:.4DE+01 4.0 1.24E+01 8.51E+OO 7.71E-01 3.'69E-01 '9.67E+OO 1.1'6E+Q1.
N E 5.0 1.02E+D1 EL77E+OO 6.42E-0'1 3.17E-01 7.7-3E+OO ft55E+OO 1.0 7.84.E+OO s .. rnE+oµ 5.26E-01 2.47E-01 5.f!BE+OO 7.87E+OO 1fW 5.04E+DO 3.35:E+DO 3.69E-01 1d31E-01 3.B8E+.OO 5.75E+OO
- Assessment Data Resul\5 S.?lted to File:_ . . M F Grand Gulf 10Miles Monitored R.el.ease 020'621J18 2.05648.UR17 H
J Reviewed By: EP-CALC-GGNS-1701 Page 38 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent SPING 7 -Alert Dose Assessment Grand Gulf Tuesday~ Februar/ 6, 2018 20:57 Method: Detailed Assessment - Monitored Ref ease Release Patl'n'iay: <RCS><Supp><Contain> <HVAC filters> <EnN> PRF: ,e,_oo E-04 Contai.nm1<.nt HllT: = < 2 Hours Containm.ent Sprays: = OFF Supp Pool S~tus: = Subcooled Safety Filters: = NJ'A HVAC Filters:= li1Jorkin9 Aux Bldg HUT: = IN/A Turbin.. Bldg HUT:= NlA RadW:aste Sidi;! HL,"T: =N.'A Source Term: Reactor Core Accident - Clad OnSii:e Lower Time P..fter SfD (hh:mm): 1:00 V\lind: frorn 270° @ 4.4 mph Release Duration {hh:mm): 1:DO E.TE (hh:mmi: [NIA] Stability Cia!:m: D Preci.pitation: None Monitor: SPING ch 7 Read.ings: 1.98E+04 ,c;pm flowmte: 6000 CfM Distance Expo~ure E:idemal lnhat.ation Deposition TED:E .. CDE Evactfation Are<'ls fronf.o. t6 1'0 Mile.s Rate Plume.DOE "CEDE Ground ODE Thyrold Circle dis.tanaes .are- 2,..5 and IO rni!,es.
*(MBes) (mRlnr) (mRem_r. (mR:em} **(n1Ren1) (mfiei,,1 .. (mEt.trri) A 1.2'1E+01 *U?E+OO .. 4.89E-01 1.00E+01 2,7_1:E+.01 R 8
... $.B,. B.39E+DO -**.--
0.5 9.92E+OO 6.84E+OO 8.BOE-01 4.U4E-4}*1 8.12E+OD 1.90E+0*1 0.7 6.76E+OO 4.64E+OO 5.ME-O*i 2.fdE-01 5S!E+OO "l.14E+0*1 1.0 4.213E+DO 2.92E+OO 3.71E-IJ1 1.96E-01 3.49E+dO S-2.BE+OO 1.5 2.39E+OO 1.62E+DO 2.13E-01 1.14E.-fi1 1.94E+OO 3.28E+OO p 2.0 *1.B8E+DO 1.28E+DO L*HE-Crl O.OOE+OO *L42E+OO 2.09E+OO 3.0 1.53E+DO 1.05E+OO O.OOE+OO OJJOE+OD 1.05.E+OO 1.4DE+OO 4.0 1.24E+OO 8.S*ff.-01 O.OOE+OO O.OOE+OO 8.51E.-01 *1.16E+OD 1.02E+OO 6.77E4Jl1 O.OOE+OQ O.OOE-rOO 6.77E-01 9.55E-D1 N E 5.0 7.0 7.84E~!H 5.18E...01 O.OOE+OO 0.0-DE+OO 5:18E-01 7.87E-01 rn.o 5.04E~Oi 3.35E-O'i O.OOE+OO O.ODE+OD 3._35:E-01 5.75E:0*1 Assessment Data Resul'.s Saved to File: M F Grand Gu.1f 10Miles Monitored Release 02D'62018 205759.UR17 K H J Reviewed By: P.a~e 1 of 3 EP-CALC-GGNS-1701 Page 39 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent AXM 3 - General Emergency Dose Assessme-nt Grand Gulf Ttiesday; February 6, io1~ :21:0-0 Method: Detaited Assessment - Monitored Release Release Pattnnay: <RCS><Supp><Contain> <H\IAC Filters> <Emr> PRF: KOOE-04 Coritainme:ntHUT,.= <2-Hours Containment Sprays: =OFF* Supp Pool Sutt.rs: = Subc,ooled SafetyFilt-ers:= NJA HVAC Filters: ='INodting Aux Bldg HUT: =WA Ttil'bin,e ~ld9 HUT: =NlA RadWaste Bldg HUT: =NIA
- Source Tenn: Reactor Core Accident - Clad OnSUe lower Time After SID (hh:mm}: 1:0U 'Nind: From 270° @,4A mph Release Duration (hh:iltm): 1:DO ETE {hh:mm): IN/A] StabiHty Ciass: D PrecipitatE_orr None Monitor: AX M ch 3 Readings.: 1.35E+OS ,cpm Ffowrate: 6000 CFM 0.5 9.92E+D2 6.84E+02 S.84E+01 S.13E+02 1.92E+O$
0.7 6.80E+02 _4.64E+02 5.88E+01 2.91E+O'l .5:52E+02 1.14E+03 1.0 4.32E+D2 2.93E+02 3.72E+01 1.96E+01 3.50E+Q2 6.32E+02 1.5 2.39E+D2 1.62E+D2 2.14E+D1 1.14E+01 1.94E+02 3.29E+02 p D* 2.0 1.88E+02 1.28E+D2 1A1E+01 7;39E+OO 1.50E+02 2.JOE;+02 3.0 1.54E+02 1c05E+02 8.95E+OQ 4.SBE+OO 1.19,E+02 1.41E-+:02 4.0 1.24E+D2 8;52E+01 7.74E+OQ 3.9"1E+OO 9.69E+O*t f.17E+02 9.62E+01 N *E 5.0 1.02E+02 6.77E+01 6A3E+OG 3.17E+OO 7.73E+d1 1~0 7.88E+D1 5.19E+01 528E+OO 2.48E+OO 5.96E+O*t 7.'9*lE+01 10.0 5.04E+01 3.36E+D1 3]0E+OO 1£1E+OO 3.89E+01 5.77E+01
*Assessment Data Resulis Sa11ed to File::*' M F Grand Gu1f10Miles Monim:red .Refeese 020,621J18 210013.UR.17 K H J
Revie11red By: EP-CALC-GGNS-1701 Page 40 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent AXM 3 - Site Area Emergency Dose Assessment Grand Guff Tuesday, February 6, 2018 21:02 f~ethod: Detailed Asses.sment - Monitored Release Release Pathway: <RCS><Supp:><Conrain> <HVAC Fflters> <Eiw> PRF: 8.00E-04 Containment HlrT, =*:C 2 Ho1Jrs Conbir,m.ent S,pr.ays: =OFF Supp Poof Status: = Subcooled S.:i.f.ei!'y Filtf:FS: = Nl'A HVAC_ filters:= War.kin:g Aux Bldg HUT: =Ni'A Turbin:e Bldg HUT:= IN.IA Racll#.aslle Bldg Hui: =NIA Source Term: Reactor Core Accident - Clad OnSite Lower Time After SID (hh:mm): 1:00 VI/ind: from 27Dg@ 4.4 mph Relea{;)e Duration {hh:mm): 1 :00 ETE (hh:mmj: [NIA] Stahilify C!ass: D Precipitation: None Monitor.: AXM ch 3 Readings: 1.35E+04 cpm flowrate: 600D CH.i1 D_ist:mce Exposure Extem:tl lrth~latiO(l* Deposition "*l'E:DE .cp~;
. -Rate *P;lumeDpE CEDE' Grt,r.md:DDE' T(lyrn1~' Circle d1:stanoes are 2. 5 and Hl miles.
(Mtles*, .. * (mRthr)' (rnRem}.***.. . (mReh::} (n"iRe:mj .*. (n~!iem) (ri,Rcm) *' A 1,0*1 f:,i-02 2_.71E+02. R 8
.. $.B. < -~
. t.21E+D2. ... BAJE+.01 .. *1.1.6E+01 ,4,89E+QO 0.5 9 ..92E+D1 6.84E+01 8.84E+OQ 4.04E+OO 8.13E+01 t.92E+02 0.7 6.80E+01 4.64E+01 5.88E+OO 2.91E+OO 5.52E+Ol l.14E+02 1.0 4.32E+01 2.83E+O*J 3.72E+OO l.96E+OO 3.5GE+O*t 6.32E+01 1.5 2.39E+01 1.62E+0*1 2.14E+OO 1.14E+OO L94E+01 3.29E+0*1 p 2.D 1.88E+01 1.28E+D1 1.41E+DO 7.39:EA}1 *J.5.0E+O'l .2.*10E+01 3.0 1.54E+01 1.D5E+01 8.95:E-01 4.58E-D1 1.1&.E+01 1.41E+01 4c0 1.24E+D1 8.52E+OO 7.74:EAH 3.91E-{li1 9-69E+OO 1.17E+01 9.62E+OO N E 5.0 1.02E+01 6.77E+OO 6A3E..;[]11 3.HE-01 7-73E+OO 7.0 7.88E+DO 5.19E+OO 5.28E-01 2.48E-DJ 5.96E+OO 7.9lE+OO 10.0 5.04E+DO 3-36E+DO 3.70:E-O*i i.61E-D1 3.89E+OO 5.77E+OO Assessment Data Results Saved to File: M F Grand Gulf 10MHes Monitored R,erase 02062018 210213.URl7 K H J
Revie~,ed By: EP-CALC-GGNS-1701 Page 41 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent AXM 3 -Alert Dose Assess-ment Grand Gulf Tuesday; February 6, 2018 21:03 Method:. Detailed *Assessment - Monitored Refe.ase Release Pathway: <RC.S><Supp><Contain>*<HVAC fJ!ters> <E1w> PRF: KOOE-04 Containment HUT:=< 2-Hours Coniair,ment Sprays:= OFF
- Supp Poof Status: = Subcooled Safety Filters: =NJA H\JAC:Flhers: =Working Aux. Bldg HUT:= WA* Tu)'b1neBldg HUT: =WA Rad\N!E!s!!i! BldgHIJT: =NIA Source Tem1: Reactor Core Accident - Clad OnSite lo*wer Time AfierS!D (hh:mm): 1:ClO Vi/ind: from 27Da@4.4 .mph Release Duration (hh:nnm): 1 :00 ETE {hh:mrh}: [NIA] :Stabill!}' Class: D Pred,pitatton: None
-Monitor: AXM ch 3. Readings: 1.3SE+03 cpm Ffowrate: 6000 CFM <
Circle dis:tanaes are 2..5 and 10 miles_ A R 8
... -~,I?, 1 :1E+O E _1. E+Ol .1 *1 ~l;:t-09 . :1-8~1:::Jij ... *. J.O'l_E+O.t .... ~Z:IJ~:1:Qt, . .
ff:5 9;92E+DO 6.84E+OO 8.84E--01 4:tME-01 :8.13E+OO 1.92E+tlt 0.7 6.80E+DO 4_64E+OO 5.88E-01 '2.91E-0*1 5:52E+OO U4E+ll1 1.0 4.32E+DO 2.93E+WJ '3.721E-01 1*.96E~0*1 3.5Il'E+:OD 6.32E-+OO 1.5 2.39E+OO 1.62E+OO 2.14E-01 1.14E"'fi1 1.94E+OO 3-29E+OO p D 2.0 1.88E+OO 1 ..28E+OO 'L41E-01 O..OOE+OO 1.42~+00 i.JOE-t:00 3.0 1.54E+OO 1.:0SE+OO O.OOE+OQ o:oOE+dO 1.. 05!=+00 1,41E;JIO. 4.0 1.24E+OO B.52E-01 0.0.0E+OO -0.00E+OO .8.52E-;o1 *L17E+OO 9.62E.:01 N E 5.0 1.021:;+00 6.77E-01 CtoOE+DO O.OOE+OO 6.77E-D1 7.0 7.BSE-01 5:.19E--01 O.OOE+OO O.OOE+OO S.18E~01 .7,91.E-:01 1{t0 5.04E-01 3.36E-0:1 {toOE+OO O.OOE+OO 3._J6E;.:01 s.nE.:.01
- Assessment Da:ta Results Saved to File: .M F Grand.Gulf 10M1les MonitDred Ref.ease 020*62018 210343.URl7 K H J
Reviewed By: EP-CALC-GGNS..:1701 Page 42 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent AXM 4 - General Emergency Dose Assessment Grnnd Gulf Method: Detailed Assessment - Monitored Release Release Patl"wvay: <RCS><Supp>"':Contain> "':HVAC Filters> <Erw> PRF: 8.00E-04 Containment HUT: =< 2-Hcurs Containment Sp1ray.;: =OFF SUJPP Poot Sk.,tus: = Sub,::-ooled Safoty Filt-ers: = NJA HVAC Filters: = 'Nor.king Aux Bldg HUT: =N/A Turbin-e Bldg HUT:= N/A Rad'Naste Bldg HUT: =NIA Source Term: Reactor Core Acctdent - Clad OnSire Lo1wer Time After SID (hh:mmj: 1:00 VI/ind: from 270°@ 4.4 mph Relea$'e Duration {hh:mm): 1:00 ETE r.nn:mm): [NIA] Stai)ilify Cfass: D Pre ci.pitation: None Monitor:: AX M cl1 4 Readings.: 7.48E+07 cprn Ffowrate: 6000 CFM Distance E::<posure *.J:>:ternat lnh.:.;latlon ;Deposition . lEDE **cor=, Rate Plume.DOE. CEDE Ground DbE .Thyroid. Circle distances are 2, 5 and ~D mil~s.
.(Mtles) .i: n1Rrtir) * (rnJ~.;n;:;* {rnRem) (111iRem;, (mRern) (mR-;m) A
**: ,:S]3~*-.* 1.2'1E+03 B.39E+02 .. 'l,.15E+02 4_.8BE+0*1 ','.,_;"! ..2.7,-1E+Cl3 R B 0.5 9.88E+02 6.84E+02 B.80E+01 4.02E+0*1 8.12E+02 1.90E+03 0.7 6.76E+02 4.64E+02 5.84E+01 2.90E+O*t 5.5-1E+02 L*14E+03 1.0 4.213E+D2 2.92E+02 3-70E+01 1.96E+01 3.49E+02 £.28E+02 1.5 2.38E+02 1-61E+02 2.12E+01 1-14E+O*t *L94E+02 3.27E+02 p D 2.0 *1 .. 88E+02 1.28E+02 *t.40E+0*1 7-35E+OU 1-49E+02 2.09E.+02 3.0 1.53E+D2 1.D5E+02 8.92E+OO 4.56E+OO 1.1HE+02 1.40E+02 4.0 1.23E+02 3.49E+01 7.71 E+OO 3.89E+OO 9a65.E+0*1 1.16E+02 6.76E+01 7.72E+O't 9-5S.E+01 N E 5.0 1.02E+D2 6.41E+OO 3.17E+OD 7.0 7.84E+D1 5 .. 18E+01 5.26E+OO 2.47E+OO 5.%E+01 7.B7E+01 10.D 5.04E+D*J 3.35E+0*1 3.69E+OO 1.60E+OO 3.8&E+o-t 5.7SE+01
;.\5.sessment Data ResullS Sa>ted to File: M F Grand Gulf 101...1iles Monitme-d P...el-ease 020{\2(}18 2.1000i.lJRl7 K H J
4.97--02 (tl0%) Reviewed B y : - - - - - - - - - - - - - - - - - -
- EP-CALC-GGNS-1701 Page 43 of 71 Revision O AA1.1, AS1.1,and AG1.1 URI Calculations Containment Vent AXM 4- Site Area Emergency Dose Assessment
*Grand Gulf Method: Detaifed Assessment - Monitored Release ReleasePattn....ay: <~CS><:Supp><Contain> <HVAC Filters:;,, <Env> PRF: KDO.E-04 Containment HUT:=< 2 Hours Contail:llment Sj:l!rays: =OFF Sup-p Poot St.z.tus: = Su'bcooled Safet1* Filt~rs: = NJA HVAC;Filters: = Working *.ii.ux Bldg HUT:= WA Turb:in,e Bldg HUT:= NlA RadWasite .Bldg Htff: =NIA Source Term: Reactor Core Accident - Clad OnSite b::wter Time After sm*(nh:mm): 1:00 Wind: fron1 270" @4A mph Release Duration {hh:rnm): 1:OQ ETE (ilh:mm): {NIA] Statiill~ Ciass~-D Precipitl,ltion~ None Monitor: AXM ch 4 R~adings: 7.48E+06 cpm, flo'.wate:* 6000 CH11 B
05 9.88E+01 6.84E+01 8.SOE+OO 4,0:ZE+OO K 12E+Q1 H!OE+02 0.7 6.76E+_01 4.64E+0.1 5.84E+OO 2.9DE+OO. 5:5"tE+01 l.14E+02' ,/ 1.0 4.28E+01 2.92E+D1 3.70E+OO 1.96E+OO 3.49E+Q*J 6.2BE+(L1 1.5 2.38E+01 1-.61E+0*1 2:12E+OO 1.14E+OO t.94E+01 3'27E+CU
- p 0 2.0 1.88E+01 1.28E+01 *h4QE+DO 7,35E-01 1.49E+O*t 2.09E+01 3.0 1.S:3E+01 1~osE+o.1 8'..92E-01 4.;56E-01 1.1aE+m 1AOE+01 4.0 1.23E+01 8.49E+OO 7.71E-01 *3.ME-01 9.65E+OO .1.16E+01 6.76E+DO 6.41E-01 3.l7E-o, 732E+OO 9.551:::+00 .N E 5.0 1.02E+01 7.0 7.84E+OO 5.18E+OO :5.26E-1;)1 2.47E-01 5.96E+OO 7.87E+Cl0 10.0 5.04E+DO 3a35E+OO 3.69E-01 1.60E-01 3.88E+OO 5s75E:+QO
.A.ssessmen.t Data Resuhs Saved to File: M F Grand Gu1f10Miles Manitored.R.e!,ease 020'62018 210713.URl7 K H J
Reviewed By: EP-CALC-GGNS-1701 Revision O
~AA1.1, AS1.1 and AG1.1 URI Calculations Containment Vent AXM 4 -Alert Dose Assessment Grand Gulf Tuesday~ f,ebmari16, 2018 21:0S Method: Detailed Assessment - Monitored Refease ReleasePathway: <RCS><:Supp><Contain> <HVAC filters> <Erv..-> PRF: KOO E-04 Containment HUT:= <2-Ha1...1rs Containment Sprays: = OFF Supp ?OIO! St.:Itus: = Suboooled Safety Fi lti:!rs:. = N.lA HV AC fib;;rs: =Working Aux Bldg !-:UT: = WA Turuin,e Bldg HUT:= NlA RadW-aste Bldg r-!IJT: = Nl ..!a.
Source Term: Reactor Core Accident - Clad OnSite Lower Time After sro (hh:mm): 1:00 V\t'ind: from 270°@ 4A mph Release Duration {hh:mm): 1:00 ETE (hhi:mm): [NlA] :Stability CEass: D Predpitatton: None Monitor: AXM cl1 4 Readings: 7A8E+05 cpm flowrate: fi-OOD CfM Distance EXt*osure -External' lnt)~lf1tion . Deposition TEDE CDE R~te Plume.ODE: CEDE _ Ground DDE ,Thyroiq Circle drstanaes are 2. -5 and tO miies. I**
,(~.1tles) . (n1Rthr) ,(mRein} (n1R'en)) (n1Re.n~1 (nif:em1 (n~F-erb) A 1.. qi:r,~+01 _2}1E+01 R 8 S.B. 1,21E+01 ,. B.39E+OO l,1?E+OO 3.&BE:-O 1_
0.5 9.88E+OO 6.84E+OO 8.BOE---01 4.{)2E-:01 R11E+OO "1.90E+fJ1 0.7 6.76E+OO 4.64E+OO 5.ME-13-1 2.90E-01 5.5-1E+OO 1.14E+0*1 1.0 4.28E+OO 2.92E+OO 3.70E-Gti 1.96E-01 3.49E+OD 6.2BE+OO 1.5 2.38E+OO L61E+OO 2.12E-Ol 1."14E-01 *J.94E+OO 3.27E+OO p D 2.0 1.88E+OO 1.28E+O_O 'i.4DE-1H O.OOE+OO 1.42E+OO 2.09E+OO 3.0 1.53E+DO 1.DSE+OO O.OOE+OO O.ODE+OO 1.0SE+OO 1.4!JE+OO 4.0 1.23E+OO 8.49E-01 U.OOE+OO O.ODE+OO 8.49E-01 1.16E+OO D.OOE+OO O.OOE+OD 6.7r3E-01 9.SSE-01 N E 5.0 1.02E+OO 6.76E--D1 7.0 7.84E-01 5.18t:~1 {tOOE+OO :D.OOE+OO 5.18:£-0*1 7.87E-01 10.0 5.04E-01 3.35E-O'l O.OOE+OO O.ODE+OO 3.JSE-01 5.75E~01 Assessment Data Results Saved to File: M F Grand Gulf 10Mi!es Monitored Rel.ease D20021J18 21082E,.URl7 K H J Revie'l!ved B y : - - - - - - - - - - - - - - - - - - - No *le Gao 6,44E+OO (99.8%) EP-CALC-GGNS-1701 Page 45 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent SPING 7- General Emergency Dose Assessment GrandGult Method: OetaEfe-d /u~ssme-nt ~ litontito,,ed Re4e~se Release*P~"ifiW:l'.f, <R.CS><Au:;.;: Btd's> < ~ W~ste><HVAC FinE<rs> <Env::- ?RF~ 3.20:1:*02 ea.rainmtH"lt HUT:: .. NtA Cont,,llnment Spr.,,!Ji$: .:: NlA S11p:,i:, Pr,d Stlw;$: ;;; Ni,A. Safety Alters.: ; NIA RVAC Fill~,: .. Worl(~ hlx Bldg HUT;:;;:"' 2 ~ 'furbifle 6ldg HIJT;.;;; ~A R<il<i','Ja'i<te eMi!i Hur: ::. < 2 Hol;lrs SDurce Te~; R~J.'!Ctof Cor~ ~~Ji;l,f'l't ~ C~d OriSlte lower Trine After $JO fhh:mrri).: CltOO Wind: Freim ~70"'
- 4.4. lllf::ih Rlease o*uration {Mh:mmj: bOO ETE (hh:l'llm); (NJA J Stabj!ity Cl?.:ts;~: 0 P~ipita:!ion; None OT rn 2.t!E+rn Hi 2ABE+o1 1,1CIE+D1 5,52E+01 5.96E+02 p 2.0 . {(;{,EJE-+-0-0: 1.-646+01* 7.D:2E+OO :2.sG2E+01 4JJOE+02 3.0 *1J':BE+01 *1,1ts+oi *1.17E+01 !:LtJ4E+:00 * :2;S4E+01 2.87E+02 4.0 1.2SE+01 a.34e:,.;oo R~8E+OO ,;t90E+OQ 2.fe;E+01 :2;2.QE:+02 E
K65E+OO 3A8HO-O 1.91E+01 2:12E+02 N 5.0 (*0~5+1)1 "'l,!HE.t-00 TO 6J:1ee+oo *t:ME+tlO 6,42E+rlil 2.42E+OO 1,32E+01 1:58E+02 U!i.O lil45+00 4.49E+(Ml 1.-56E+O{J t3.49E+OO 1.f2E+02 M F K H J Reviewed. By; Gram:!: Gillf 12.0:1.0 EP-CALC-GGNS-1701 Page 46 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent SPING 7 - Site Area Emergency Dose Assessment Grand Gutt Method: Oet.aEl~d As.'5.e,ssmf!-rlt ~ Mie,nlt~red Rclc."\ill~ R-elease P,;1~llww,.: ,;:J~C.$:::> ',,,Au11. 810;> ~,R~ 'N,iJ,st;a,> <HVAC Fit!t7rs:,, <:Em;;" PRF: ~*.20E,02 C.or,;rainmi;l'!t HUT: ;:; N..l,. Cc;,nt:i!lnmenr, Spr.1~$: :: N'A $lipp Po:;~ $1:at*.;is: = NJ.,!; $::ifi;t.y Filters: "' NIA HVAC Fiil.;:r:a. ;:; 'Norit:r,g A.I.ix, Bldg HUT: *= ,( 2 Ho,."fS Turbine Bldg HIJT;"" t'll'A Rad 1.'Ja,;:te Bt<f*g HUf: =*,; ;2 HOi.lrs Source Terrii: R~.::i,ttor CQte;, f!.i."li~j~~ - Cka.d OnSiie tower Time Afte:r $.'C.I fhh:n-i1r1}: OJJ:{J v~r;.nd: Frt:iff1 .:aD~ *@4.4 rni:::;h ITTE (hh:mi:n): (NJA J SUbl!ily Ck1s~: 0 PrJ?Cipita,ton: Ni:::<r1e-Rea.-dings: 1.72E+iJ3 cpm lr!n.llla~i6r(
- TEDE**. CDE_
CEDE "Gn:iur,d .[}'.J'E Tn*)*roh:f (Miles) 1n1Jtlhr(, (rnRf.!rn.;, ,;rnR1?t11) ,:r{;P,:em) frnR1::tt1) A S.S. .\riE:+-01 1.99E+01. _B.34E+Clfl_ 3.~E+0.1_.. _ .. 5.DD.Et02 R B D..5 1 Bc,E+Oi 1.40E+01 tlD4E+O{J 2.aGE+01 3.51E+02 0.7 !teae+r:H:i 8.3r3E+t1[J 3.72E+IJD 1.78E+01 :to8E+02 HI 5.-._..,...., -~ 4.64E+D:ll 2. l tE+OD *1.[HE+0*1 1,l4E+02
!L5 to,::,E+ilb 2A5E+0{1 1.10E+[l,0 5:52E+OO 5.Qe;E+01 p 0 2.{l ~-EISE.,(U 1J.}4E+Ofl 7JJ2E-01 2:92E+0;) 4.IJOE+01 3.{I 1.tiBE+OO 1.1 iE+OO fLD4E-OK 2.84E+OD 2.87E+01 4.0 1.2.Sl:2+0;0 Q.3SE-IJ'F. 3.9-iJE-IJ.1 2. lBE+C(l 2.:29E+01 5.fl 1A)~E+l.'}0 B..6.5E-IJ.t 3-ABE-Ol 1.ftiiE+OD 2.12E+01 N E 7.{J \}.@(1EA.H 6A!2E-Dt. 2A2E-{H 1.32E+OO 1.58E+01 8.4'9E-t:H 1.12E+01 Assess.ment 0*,;n~ Ri\:~'IS. S.'!'it-:J to f\'.\f;. M F Grand Gulf 10Miles. M1lt1it<<ed Reli~,; G40~l'.!17 OBD.142.URll K H J
Partic1.fatB '9.73E-.f!3 ('D.2%} lodi:ne 2.3~E-Ot (4.6!'-:to) Revievited B y : ~ - - - - - - - - - - - - - - - - Ncble-Gas 4~90E+OD (9.5 ..2%} G."3,1ll Gulf J 2.0:1.D EP-CALC-GGNS-1701 Page 47 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations RadwasteVent SPING 7 -Alert Dose Ass,essment *- Grand Gut, *rhun:id3!)'; Apria s,, iOH'. oa:02 Mettlodl~ Detailed Ju,s,essment ~* f;{!lowtored Relel!lse Release Path\VilY, <R,CS::i> ~Ii;!.;;)!: EU~> <Rad W&$te':>> <HVAC Flt~f$~.. <Env:,.. . PRF: ::!,,!2PE~02. Coo,rainm~nt HVl,':
- Ni\A Cor1~!nm.snt .$pr.ii~: .;:; ~iJ.tJ!. ww Pix:~ St;r1fl'.,glii; =:NiA. $1;1fl;ey F~~ri: ;:: NlA .
tfllAC F~~;. i. Worlt)!lt /!>Jx 91ctg HUT; ;;: < 2 Kct>urs Turbin~ 61dg Ht.IT: .. f'i'A -Ra,r;J 1//.;1$,~ 8ildg HVf'>;;:;; *.:c: 2 HQi.lrJ Source Term: R~.aetm' Core-A~l!;fl't - C!'acl OnSite lower
- Time After $.tO 1'.hh:rnm}:; CtOO 1/.iind': From 27(1,;, @ 4.4 rni:*h Release- Our<ltlon (t'th:mm)~ 1;(10 ETE {hh;mrnj; (NJA 1 $t;;ibility C!a$i, 0 Precipita~ion; Noro~*
Mon.10r: SPING eh 7
-0.1 J. 72E-(H 1.::?SE+OO
. *1.DtE+OO, 1A4E+01 1.5 5.96E+OO p D 2.0 4::00E+Ct!l 3.0 !LOOE+O{I, 4.tl o.aoE+o-n O,OOE+OO O.OtJE+O:J
{LOOE+OO O.OOE+OO N E 5JJ O.OOE+OO 7.{J O.DOE+oo {tOOE+oO O.OOE+OO 1~5-SE+O!J. 1D-O O.DO.E+Ofl {l.OOE+ll!l 0.00E+OO 1.12E+O{l Assess."Uent D.1~ ,e~iS. $av to F~,;;,; . M F Grand Gulf *1 OMiles Maffitered R~~-41Sit ~017 'DB02S.t .URl1: Reviewed By;, Gr.ai".141 Gulf t 2.0.' .0 EP-CALC-GGNS-1701 Page 48 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 3- General Emergency Dose Assessment Grand Gutf Mettiodl: Oet.lFl~d AS~'iSrillil}f'i:t ~ ll!lonitl)t>ed RecS~~S+e Release P~tnw;;:;~ <::RCS::~ .~Au.i,; 61(.ig::;. <Rad *N~sW* <H\tAC FiftiS'rnt>- ...::Erw> PHF: ;,t20E*02 C-011,rainn1,;,r1t Ht.IT:. .. Ni'A Conb!rt1'!"J;,flt Spr~:f;,: *::: Ni.A S'!FW P~ $t.at*J~: ::: NJA $al'~fy Filt;,:rs; ;: NlA H'iACFi'i1~s ;:; WC>i1i:lf:,1Jl Al,.1i Bldg HUT:;: 2 Hr.>i!fS iwtine Sldtl HLJT:.:: 1\1:A RJd\N.2A'tE'. El1dg HL.!r ., "' ;2 Hours. Source Terrr:,: R~:.'lctcr C0tft !lv~ern
- CiSid OnSite- low.,.r Time Aft.er SlC'.I 1'.hh:mm}: CkOO V!11.nd: Frc:,rn 270"' @ 4.4 111;;::h R-elease OiJr,lti,on {hh:r,1,rn;i: i;{JC ETE (hh:n,mJ; [NJA] sutrnt!f' Ck~$.s: o PrecipiP.J~ion: None Monitor: AXM ch 3 Ft1:;adiogs: i. 17E+03 cpn,
. COE*
*in:rr,jid
{,mRiim) A R B 0.'5 1.40E+02 3.5\E+0'.3 0.7 8.35E+D1 3.72E+D1 1.78E+02 '.2.0SE+03 HI 4.e4E+G1 2. 1'fiE+01 1.D~E+02 1,14E+0:3 l.'5 . 'i.~'l~+Oi 2.45E+01 1.10E+D1 5.52E+01 5.9i3E+02 p 2.0 5.86t!+Otl 1.64E+ll1 7.02E:+t10 2.Gi':2E+01 4.DOE+02 3.(1 L 1~1:+tfr 1.17E+ll1 5.IJ4E+OO 2.S4E+01 2.87E+02 4.(1 1.28E+01 6.3-4!:*0L"l 9.38E+frD 3.BOE+O{l .2.16E+01 229E+02 5.0 V0~1E+cl1 7'.!HE+Oi.1 a.ec,E+ti{] :3.48E+0,{] UHE+OJ 2.12E+02 E 7.0 (l.*(lfiE+OD 4.34-E+O:(l 6.42E+f!il 2:42E+Ofl 1.32E+01 1.5'8E+02 1lD.O J,64S+CO 2A:;;S+0:{!1 4.49E+l}ll 1.!58E+OO .B.4GE+OO 1.l:2E+02 Assessment D.:au R,;$<.;l:iis S;ru~ tci fit~. M F Grand Gulf 1OMl!e~ Mi:mite,ed .Reli;<,;ii~ t:14ilil:2011 07552.5.URl7 K H J Partir.:uat~ ft72E-D:2 (D.2%} lodfn~ 2.38E+OD (4.5%) Revie1-"v"Bd B y : . ~ - - - - - - - - - - - - - - - - - Noble-Gas 4.90E+OJ (!;l5 ..2%} G.'31::le Gu!1' I' 2.0.1.0 EP-CALC-GGNS-1701 Page 49 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 3 - Site Area Emergency Dose Assessment GrandGtdf ReleaseP~~liwlly:: ~;RCS:.,. <Aui 9.J~> ~Ra(J W.i3:$W'llr<HVAC Fiheti> <Env> PRF; :l.20E~Ol C.Onta1nmerit H\lt
- NIA Corit.ilhJ'f!i;lflt 'Spr;:11!,$: *;;;: NlA ~p Pw $t.,ti.,.s;.: NtA S;afety f ilti;;ri; :;: Nh\
_HIJAC Fi'lltr$;.;;; Wort~~ A>..1:r.: Bldg HUT;;;;:,<: 2 ~ T!Jrbine Sldg HUT: ... Nf.A Rai:l'I/~~ BM.g HI.Jr;.;:; *<:. i HQ1,.trs. Source T~rm: Rie-.actor Ct:tfi:- ~eni: - C&Jid On:Si,e lower Time After $,10 (hh;mt't* 1);00 . i/iflnd: Frorn 270"'@4'.4 mph Release.Oi,,wo.lion(hh:mm); b!lO ETE{hh:mm); (NlA) SUbHity Cliii$$;: 0 Pra(:i pir.ation: Noris-B 1AOE+ll1 6. 04E+O*O 8.3'3E+0{1 *3.72E+mJ 2..08E+02 4.64-E+OO 2. UE+O.Q 1:CHE+01 1,14E+02 2A5E+OD . *1.}t1E+(Hl 5:52E+OO 5.96E+61 p 1.5 0 2.0 1;ME+OO 7,02E.:DJ 2~Q2E+OO 4.00E+.01 3.0 2,84E+O!l 2.87E+01 4.0 2.l6E+O!J 22ge:+01 2:1:2E+01 N E 3A8E-Dl 1.91\E+OO 2A2E-01 1s32E+OO Ui8E+01 8A:9E-'01 1.f2E+01 Assess.,'l1ent o~~ .e~i'JS, s~, io f* <\;'.; M F Grand Gulf 1OMile~ M<mit,f>ff:d R~&J/5,1;'. (l4400;.;0l71Ji"57~.lJRl7 K H J Revie"W'ed By: ..
... Gra:u1 &JI!' J 2.CL1 .0 EP-CALC-GGNS-1701 Page 50 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 3 - Alert Dose Asse:ssment, Grand Gutt Method: Oec.;:i.Hed A!i$e,~sme:nt ~ ,Mi;i.nltc-e,ed Ri?.Jf;>~;S,e Rel8ase P;;11tiw;3:y: ~;RCS::~ ,:'AIJ:;;;. 13ldf.f'* <F?.zd 'l1iast1;;;. <HVAC FiltE'rs.:"' <:Env> ?HF; :3;,.iOE-02 Cooirainm,;,rtt HVT: ; ; N<'A Cant~fnmenr, Spr.1!~: "' NfP. $w:!)p Pi):J S!:.11;'.ns: :::: NIA S.irety Filters; -= NIA HVAC F~1~$- ,;:; V*.rorl;;,r,g Au~ Elldg i,,iUT; "",,;; 2 Hours T1Jrbiri1: 81<:1f1 HIJT:. :::: Toi.'A R.71.d'.'last.E: Btd11 Hlff: :::: ~ 2 Hoo rs Sourc1= Ts,tm: R~.~et()r C0t~ ~,tt,:)~~; ~ C&~d OnSite lower Time.A:fi:1:r $.10 fhh:rnrri}: C,00 il~'i.nd: Frcrn 27fl<I< 4.-4 rnp::.h R'E:le:ase Dvralion ( hh;mm):. l ;{JO ETE 1.hh :rnrn): (NlA ) St;;;bi litf Cl)ass: D Preci pjr.aiicn:: Nc,r:11:
Monitor: .AXM <;tl :3 Rea.dings: 1. 17E+D 1 cpm flrYwrat,e: 5249'5 CFM
- Dist:m-re Ezpf.Y:i,..i.rE fi::t~f!:ffi~l 'ir;riala,ion' Clei:,'<lsi~ion . TE.DE. COE R..'l~E F-iUJ:lls:' GEDE Grt,ur:d: DD E Th*>*roict (f'~ti!E:S] \01P:/hr:* t1~\R~110:; tmRem~ ~n1-R4>nv;t_ (.rhRevn), (rnRe:1111 A s:s. ,7:;f;i;+OO . 1.J:;E+Orl, , 1.99E+!Jfl .. 834E-O~ 3.Q4E+OD 5.00~+0J.
R B o..5 1.3,e',£;+110 ftME-0 1AlOE+OD llJME-0~ 2.S@E+OO 3..6\E+01 0.7 S.l:l8E**O~ 5.72E-Cl fL36E-Ot 3'.72E-O~ 1.78E+OO :L!i3E+01 HJ 5.. 2{1E~IJ! $.3.l~AH 4J34E-Di 2.. 11E-G~ 1.DhE+OO 1,14E+Oi rn 3:ofiet-or. !.J.l"fE-6~ 2.46E-(H f.lClE-D; 5.52E-CH 5JJf.E+O!l p 0 2.0 0,001::+l'.NJ ,QJ!OE+D-0 1.ME-[H 0.00E+OO 1.e4E-Di 4:00E+OD 3.'D 1. 7.\il'E**Oi l,t,1:1~:-CH U7E-!H O.OOE+O{l 2.33Ealll 2J37E+OD 4.{l 1.2.ae-oz ffOOE+O{l O.OOE+IJ{l Q.110E+Ct!:l 2..1GE+OD 5)[1 l.O{JE~(H *OJJ1JE+OO {UJOE+Dfl tl.OOE+O*D O.IJOE+OO 2.12E+OO N E 7.{J fJ.l!OE+tJO D:0)E+-t:l0 O.OOE+[J.fl fl.DOE+O{l O.OOE+OD 1.58E+OD m.o o:oc1e+c:0 fi.'l'JOE+o.o, 'D.OOE+fJD fl.OCJE+{l,!J O.lJOE+OCl 1.12E+OD A.s:sessmertt D.jjo Ri:sd:ts. S.,11,11'.';d 1t:I fb:s,. M F Grand Gulf 1OMil1t:1> Mor1it,ererl R~tP.'7#S~ l'.t410~::ZO l7 075tt2'.,E,.URl7 K H J Partictfate {l.72E-ID4 (.02%} lodfni: 2.38E-02 (4.6"-'5ii} Revie*wed B y : - - ~ - - - - - - - - - - - - - - - - - - Noble Gas 4.90E-O t (:05.2%) G:an:!, Gull' i' :2.0.1.0 EP-CALC-G_G NS-1701 Page 51 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 4- General. Em~ergency Dose Ass,essment GrandGutf Method: OelaEl(ll<tAsN,.,sment =.f.1lontt0:red R~e.ll~ Release PJ~l'lw;.1~: <RC.S> <Aui Eild,> <~ *N.;llstti> <HVAC fll*ief~> ~ienv:,,. f'RF: 3.20,E~O:! G.cmrainr~nt Ht..Jf; .. NA:.; Cont.Jh'lf'!!l;flt Spr,11y'S: ;; NJJI, Silrw P~ Sta~s: "' .NIA $~fety filter$.; ;;:; WA HIJAC Fi1en,,'.;;; WQ\:k~ NJx Sldg HUT;;:: ;t~:2 Hollr~ Turbine Sldg HUT:;;::; NIA R~4\'ilas~ Ell<i'-g! HUT: ;:;; *..: :2 Hot.lrs. Source Tern,: R~4.'leter C~1;- ~~d,!;fl'i: ~ C~d OnSliea....owe-r Time Afte.r SiO {hn:mm}: O:;O<! IJ'l!ind:_FrQrf1 270"',@ 4.4 rt1J:ih Release 01;;1r,3ti,Qr,. 1;{10 Era {hh:mrn_~,: (NJAJ Stability Cl~i~: 0 Preefpi~ation: Morre Monitor: AXM (Ii, 4 D.5 f.3:5E+O::? CG a.aae:+c1 HI 5,:20E+01 4.64E+cn 2.HE+01 UHE+D.2 lA4E+03 Hi 3:.tie,e+o1 1Ji-1e+of .2.~5E+01 1.10E+Di 5J:i2E+01 5.00E+02 p 2.0 {U:l!E+OO 5.SO£i{MJ 1.64E+D1 7.02E+Dfl 2.92E+0'1 4:00E+Q2 3.0 1.1se~o1 1J~E+il1 1.17E+fl1 5JJ3E+Q.{J. :2.83E+01 2.87E+02 4.0 i.2ae+o1 tll.:34E+OO .9.38E+OO. 3.90E+OO 2)6E+01 ::.t29E;+02 5.(1 tO~E,+01 1~0,e+orf .£Lfl5E+O{l 3.48E+O{] 1,91Et01 2.12E+02 N E 7J) fJ:{leE+OO 4,34E1i-OO 6.42E+OO 2.42E+O{l L32E+01 L58E+02* 1.~fiE+Or.1 B.49E+OO 1.12E+02 Assessment ~R,;i~ S;;ri1edw f~; M F Grand Gulf 1OMiles. Mofilt(fe-,:fRJJ;41;'~~06;."(I 17 07'585e,URl7
*H J
Reviewed By: Grarui Glllf I :2.0.1.0 EP-CALC-GGNS-1701 Page*52 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 4 - Site Area Emergency Grand Gutt lti'lettlod: D~ta.Eleid Assessme-nt
- Mol"lli'ti:>t'ed Reh:,~~~e Release P::i~hv.{;.:r -~:RCS~"' *1:AUlol Blrl:i;i:~ <R~ 1/1/qiste,;:;, ~-:HVAC FiH<i;-r:s:" <Eriv> PR:F; ;.!;.2QE*02 Con,ra:inm,;;r1t H1:/fi: ::: NiA Con~lnment Spr.;i~: =NlA $iipp P~,, 5t,.,t,.;;s: ::: ~lJA $;,sfety Filters; "" NJA K1.I.A.C F~1,;,r'$ ::: W(lf'll,rril! AJ.1x Bldg HUT: ::: .;;. 2 ~-*r:t_1£rs_.___________T_u_rb_in...e_8_l_dg~*_H_U_T_;,.._-_N_*A_._ _ _ _ _ _ _ _ _R__a_d1_.*,_2a_":te_:_B_ta.,...,r;;_~_*FJT_'_:::_*"'-*:..._;_H_cu_r_s_ _
Sour,::,e Term: Re.l;fc~r OOJ'e- ~i;:J;;-n'l. ~ Ct:.d Orn$ir.e- ii_.,,Ner Trme Af'<l:'r $10 fhh:mm}: !'.UH'.! !/find: Frc,m 27[1"'@4.4 rni;::!1 ETE {hh:mrn): (NJA j $Ubihtf Class: D Predpi~.:.!ion:: None-Dist:m,re .... .f::;i;HHnji
.... D;:;,twsi~ion TEDE CDE A;a~e- f';l.J,~).,? CEDE. .f3rc Uf:d>i:JDE 1 Tiwroid Circle .:i:st:ao~s are :2: 5 aA4 10 1n,:-2s.
i)f'.,1ile;;;:, ,;r)1P.:)hi;i (~f';8;*:r:J,";), (r:riR-i;nt} -1mRen1:) JrnReui,:i (roR:1:n,1) A R B SB. 1.7.t'.'.tE+01 1.1:?E+Qi_l 1.9£.il::+1}1 8-3:<lE.+O:p_ 3.G4E+01 ., _5.(l0,_Et02 0.5 1 35E+Oi i;LSOE+OlJ 1.40E+Oi1 tl04E+IJ{J 2.8GE+01 3.5lE+02 0.7 a..ese+on '!i?'le;+C,µ:J 8.313E+t}O 3.nE+OO 1.78E+01 2.08E+02 Hl 5.205+110 3.30E+On 4.134E+IJD 2.'I tE+O{I UHE+01 1,14E+02 1.5 toe,e+hb (~?E+OD 2.45E+O{i 1.'IOE+Ofl 5.52E+CXl -5.95E+01 p 2.0 fH!,ElliH1'l ij,1:J;OEA'lll 1.e4E+0{] Hl<2E-Ol .2.92E+OD 4JJOE+01 3.ll) 1,7,E:E+OO 1.115E+Otl 1.17E+IJO 5.D-3E-Ofil 2.&3E+OO :2.87E+Oi 4.(1 1.2s12+00 tf!.J~g.i,, 9-:.3.BE-[H 3.. 9:0E-ln 2.*15E+OO 2.29E+01 13..flcE-Dt 3.4BE-Ot 1.91iE+OD 2.12E+01 N E 5.0 1,0QE+t'.HJ 7-0HUH 7.{I e,_g:flEHH 4.34ita-On llA2E-Dt 2A2E-Oi 1,32E+OO 1.5*SE+01 rn.o Zl-Jl4E,OW :2.4:36,-0m 4.4{JE-D't l.56E-D1 8.4'9E-m 1.12E+01 Assessment O~t.1 lle,s.,.itl:5s $~-.r:ci to \?~ M F Grand Gulf10Mile;i. Mimit1:,rt1d R~l!:;ty,Wii;; 1);00:1f.ll7 07592WRl7 K H J
**J Iodine- 2.38E-Ot (4.6%)
Reviewed Sy:_~----~------------ Noble Gas 4.8ttE+OO (95.2%}
*~"am! Gaff 12.0.1.0 EP-CALC-GGNS-1701 Page 53 of 71 Revision O
. AA1.1, AS1.1 and AG1.1 URI Calculations Radwaste Vent AXM 4 - Alert Dose Assessment Grand Gutt Thursday,Aprif 6., 2011'. 07:59 Method: Detailed A$'5-e'55me-rit ~ Mool'tor~d Re!f;;1i~
*Release Pathway: o.tftCs:-.. -tAU)f. 9ld9:,; <~ W~s'!ifr, .;:HVAC FWti;;.r~)' <Eriv::.: PRF~ 3:lOE;:02 Coo,rainrJ11;trtt HVT~ ;;; :Ni'A Cc:irit,,linm:ent Spr.,"j'y$; .. N.rA wfp P~ St.lt<.As: ;:; N1A $;,ifet.y Fil~r~: *"' NlA HVAC Fi'll~: .. Woni~:.r>9 hi'!(. Bldg HUT;.;;:< 2 ~1.1tS Turbin!l 6tdg HUT!;;;: Jli<A Ra.dW;;i.,ste &49 HUT; ;:; ..; :2Hai.1rs Source Term; Re-;ebor Cor~ ~~ld-~i - Ct.J.d OnSlte lower Time After.$10 fhh:rnm}: CtOO V~tnd; From 270"@4.4 mph Release Duration, {hh:ffl.rn}; 1;/1'.IO ETE. {hh:mm): (N,'A J Stability Cl.as$: O Pr~i pha!ion: Nor:,e Monitor: AXM e;h 4 Fl~te; 52495 CFM Distil~*"* * ., l'.=xp6:it:1.f$ l;c;_.)3t~e-rn , thnalaiibri' Cilepasitfo.n Tl=cDE;.::. COE. * :};E;v~cuatfonAr~asf/oin (fto JOJ,Mles:.
.'T_h)'f8,a 0
. Ra~e, .:Pl,rm;:. l'V'li::;. .: CEOE p'ro4f;t;1)D::JE ,.*
- Cirde&stano.:s are :2: 5aJtii 10 mtt-S.
~'~*1,.~~--"~:1:.~:.:--~~*.t*{**~}~.* j,'~,(ni~,Ri~'tt.'.h~A~,:-i:'.J*Sf*:~***~*-~~--~~i.J::. :':1cfnQJ:l!J:R~.:~:.'Jm~i~~~Jl.h:JJfi~R:!;'i;~n1~1)~~~:'*~k~~iR~im~)"~4*.~*(~.ff~,R:~:e;'~~\)L;,::..i, R' . A 8
$ :a. PPE+G"!l,.. _J.J 2EtPll J 1Q9E:+Cl'.0 :0.,4:il;:-Dt, __ .. 3. ME;tOO,. . ..5.:WstPt 0.5 1.35E+a-o B.1lOEA1! 1.40E+Cl-!J fi_D4E"'CH 2: 8@E+OO 3.!:HE+01 0.7 S.fi6E~CH 5.72E-t:H 8.3.flE-Gt 3-71E-D1 1.,78E+OO 2.08E+01 5".:2-0E~rn :l'.3:0E-0:1 4L64E-Dt :L11E-Oi UHE+OO 1..14E+01 1.:5 *ia,6a.or. \.G7E..O! 2.~E-Dt 5.52E-Ot p 2.0 O:OOE:+-l'J:-0 !LOOE+OO L64E:.CH O.OOE+O-O 4JJOE+OO 3.0 L78E~Ot i::i,eeAit 1.17E-OS O.OOE+Ofl 2~87E+DO 4.0 L:JflE-0~ {l;OOE+o.o O.OOE+OO O.OOE+OO O.OOE+OO 2;29E+OO O.IJOE+OO ,O.OOE+DO O.OOE+OO 2.12E+OD N E 5.(1 1.0ilE'.~Ot '1lOt::1E+OO 7.(1 O.OOE+OO ,0,,00E-+{Hl' O.DOE+OO O.ClOE+Clfl O.OOE+OO 1:!58E+OD m.o o. ooe+co O,-OOE+-00 {LIJOE+OO O.OOE+IJJ.l O.DOE+OO 1.l2E+OO
*Assessment 0-:i~ Re~ $i,ll,!l;cl to f*~- M F Grand Gulf 10Mi~~ Monite<ed R*3se.~002017 D'l594fUJRl7 H
J
- 2.38&02 (4J}%}
Reviewed By; . 4J:!QE-01 ({15.2%) Gram:! Gutr 12.0.1.0 EP-CALC-GGNS-1701 Page 54 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent SPING 7 - General Emergency Dose Assessment Grand Gutf Wedn-s.i:i.:liy, Aprlt 6, 2017 14 :40 Mettlod: ;Q.ei.1ilt?d A~sessm~nt - Mir.mitor£-d Ref11;~s,e Release Pa~h',\lfj'.)i'. .::HC.S-"' "'-'."fS Eltcig> ,,;:hVfl,C Fil~t;1T!l?* <Ers._,:~ ?FW 8,00E,02 C.o:iitainmi;,nt Ht.}1': ;:;; .l*J.'A Containment Spr;;a~';,: =N."A $-,:pp P~,1 $t..1t*,.s: =N1A S.;:ifety Filteri; ;= NJ/;, HVAC Fi'.?!ers:;;; Wono.Jkgi Al.Ix, Bldg HUT:
- NJA Turbin,;, 81tl[I HUT: =<'.:! l'-i,ou'I';; Ra.dW:;sti; B;<a~ Ht,.JT: =NIA
.Source TE-11*1::.: R,;-::iicr.ar Core .~::l.a:i1~- - CL:.d OnGiie- Low,;;.r Timi:Af:e-r $,ti:) 1'hh:rnm}: UHl 1,Wnd: From :27[) 0 ,@4.4 n,ph R-ele.ase- D1.1r.;,ti,on.,;hh;n~rnt U!O ETE (hh:m1r1): (N.'A j $t.3ll;;i !,tf Ct,,ss: 0 Preci pir..:rnon: t-.i'.or:1e-Monitor. .SPING eh 7 Dist:.:in*.:>: E:.:"pc-suo:: E:.(:frrnj\ !n'r)~!a!ion [:~posi~ion TEDE CDE , ,EvacLtat,6i1t Ar£'.as Front Oto 1Q Miie-s.
Ra~e.- F:u:r;~"" C*D6 CEDE Gn:iuc,d DJE Th*~*roid,
,:Mi1s'.* J~1.1'q'.:r,r:, CeJ!\F-l:i;r,,;i (,mR£>m;, ~n1,P,;er11,) (n-1Re1,v,) (mREr'r;) A S~B:,. 3.7:8E+01 .... :'.!4E.-Cr 1.'fl8E+ll2 7.70E+1}1 2.9?E+02 ::;!;{D1;Et03: R B 0.5 3.0eiE..j..c.1 l.S4E+01 1AtE+G2 5.42E+D1 2. l4E+02 3.5SE+03 0.7 2.D6EH}1 1.2ei:,E+Ct1 8.5t3E+01 3.31E+D1 1.31E+02 1..16E+03 HI 1.2~1E+D1 7.aoE+o.a 4.80E+8'1 1.85E+G1 7.44E+01 1.2:E+03 1.5 1.1:if:;'+Clb 4.44E-t>C:u 2.!:BE+D1 RIJOE+O{J 3.9'2E+01 t(32E+02 p 2.0 5 . 1131=+0,[l 3;ii'.'.IEH1l1 1.57E+01 5.91E+O{I 2.49E+01 3.94E+02 3.0 ~.{15E+CO 2..t'f3E+Cl!'.l R76E+OD 3.B2E+O{l 1.60E+01 2..-4t3E+02 4.{l =i.3:E+l'.J-0 2.11E+C.fl S.05E+Cl-O 2.ME+CD 1.3iE+01 2.IJ3E+02 5.(1 2.B:;,E+C-0 179E-,..(Hl 7.IJ2,E+O{l 2.53E+OD 1.14E+01 1.77E+02 N E 7.0 '.HJRE+CO L25E1-l'.lO .5.30E+[JD 1.86E+ClD 8.4nE+OD 1.33E+02 li[}.0 1.35E+,C{I il.79E-CH 4. tOE+DD 1.39E+OD fi.37E+OD 1.03E+02 Assessment O.;it3i R&~.::i"lS 8;:;,;.1i;i:J m ft':. M F Gr.:ind Guff 1OMi!~!i- Mr.:inite<>=<i Ret.;,aSi' C-'4052017 14402:,.UfUl K H J
PAGs Exceeded in Design.ated Areas
./ .
.Re-lease Rates. :(Ci .I sec}*
Partkafat-e :B.9'2E-£12 {D.7%} lodi.ne- 1.72E+OD (t2.8%} Reviewed By: _ _ _ _~ - - - - - - - - - - - - - Ncble '3as. 1.16E+O i (86. 5-*;'D} G.731::! Guff 12.D.1.0 EP-CALC-GGNS-1701 Page 55 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent SPING 7 - Site Area Emergency Dose Assessment -- Grand Gutt W~i:JoesdJhy, Apnt 5~ 2017 14:40 Method: :Detailed Aa5essment - flooltored R~e~s,e Release. Pijihw:;\y: ~C$)* 48 fff4;:;;,. <:HVACFil~l;i:'5> <Enw~ PRF: 8.005-02 Coo'lain~t Hltf: ... WA Contairui!.enfSp~!f$; = NlA $,,pp PQ$ $1:.;l~.,r,li: :;:: NJ.I.. S:fifet,t filters: -;;, N'A Ht/AC Fi'lers:;; Wen,.~ AJJx Bldg HUT; =MA Turbin~ 81dg HUT;= <2 ~ffi R.JdV1'~$ti; at<l;g H!,,.ff~ ;; NIA OnSlte Lo*,oJ,s-r _Time After $,10 fhh:rom}: l ~C'-0 'li!ind: Frori'l 270Q ,@4.4 nlf!;h Release Our~on (hh:rnm); 1;()0 Era. (hh:mm); [N.'~] Subllity Cf."!$~: 0 Pre<:ipita;lon: None-Monitor: SPI N:G eh 7 5.48E+O,O 2.14E+01 ~~5BE+02 0.7 2.06E+O-l:l 1,26E.+1i'O 8.56E+rnJ 3.3tE+IHJ 1.,3liE+01 2,16E+02 HI 1.2~E;+O-O 7,G6E.*O-; 4.80E+OO '1.85E+O;[J 7.44E+OO i.2JE+02 B.60E-D1 3,92E+DIJ 6.32E:+01 p [) 5.81E-Dl 2t~~E+OD 3.B4E+01 3.82E-01 1.f!OE+OO 2Ai3E+01 2-JJ4E-Ql t3tE+((I 2J}3E+01 2.53E-Ot *1.14E+O-:J L77E+01 N E
'l.86E-O~ 8A1E-U1 1.33E+01 1.3'9E-fn 5.4'9E-01 1.00:E-f:01 Assess.."l'lent 0.JU R::<=..illS"!s. $av. j to Fil~. M F Grand Gulf 1OMilea Manit<<ed Rel~~ ()gOfiiO 17 144056.URl7 K H J
Partic1.fate K92E4l3 (0~7%} lodfu~ -1.72E-m (12.8%) Reviewed By: Ncble- Gas 1.16E+OD (88. 5%} EP-CALC-GGNS-1701 Page 56 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent SPING 7 -Alert Dose Assessment Grand Gutt W1:dnE<sd,:i.y 1 Apr.it !5, 2017 U:4*1 f.'letllod: Det.atled As<;;.essrnent- M:onltored Re1,e.as.e R~lease Patti.,.._,.:({' *:f<C s:;. ~:rs 8b:lfr-"' *..;H',J,t..C Fil~~s> .,:Eew> :PRF* .S.On:E>02 Oon,tainr11qflt Ht/F: ;:: N,'A C-ontainmenr. Spr.11y'$: = N!A $u,pp Pc-~ 5t.::Jt11:s: = ~.IJ.>\ Safety Filt~rs; .:: Nl/1, H*.fAC Ft11i;rs: ,;; 'Norlcr:"9 lv,ix f:lldg HUT:.;: N.!A = Turtini: Bldfl HUT: -c 2 P"1u~!5 RadVilastE= ffid*g HVT: ::; hli'A
.Se<urce Term: R,;,:;rict;,or C~\\!' ,i;..ockl,~*rn - Cud OnSi~e- low!frr Time Aft.er $.iO (hh:rnrn): LC-0 v~*\.nd: Frc,m 27D~ ,@4.4 mph ETE (hh:mm): (l\L'.A J $tabi !1tt CL..,s;.: D Preci pi:a1ion: t*.£or1e M.onitor: .SPIN:l'.;i ch 7 Readings: -l-.2e,E+02 cpn, Fiowrate: 51}DD CFM Dist3n*.:>: Exp,Js;.;:r~* E.-:~::r:-rn...ii! .lnh.i!a~ion Deposiiion TEDE COE E11ac'uat,ot1 Area~ From Oto 10 Miles R~1~1:: P:i1i:t1.o:- DDE CEDE Gt"Our,::fDJE T.h'>'rr:::iid
,;Mili;si l,r)1PJhr) ,:f111,P.,*?J\'* './ (mR,,en~} (111:R.-et1i,) (rhR~m) (rnRi::11}) A 3.B. 3.. 78E-O" 11.JiE..l'.}1 1.9::,E+GD 7,7,0E-p+*~ 2.!}7E+OO 5.02E+01 R B 0.5 :3-.05E-D! LB4EA1, 1A\E+OD 5.48E-DJ 2.14E+C(J 3.5BE+01 0.7 2.0aE-Oli !.2dE-O, 8.E*5E-G: 3.31E-CH 1.3~E+OD 2.15E+01 HI t.i9E*O~ ,!;lOOE+O.O 4.BOE-D~ L8-5E-Dl 6.t:,5E-IH L2:E+01 1.5 0,{JO~+(j{l o.ticE.;co 2 . 51E-Di O.OOE+[l{) 2.51E-D1i 6.32E+OO p 0 2.0 O.OOE+OO .tJ OC:1£+11-D 1.57E-n: ll.OOE+OD l.57E-D! 3.ME+C.IQ 3.0 O.OOE+C,O {.lAJCtE+C-0, fl.QOE+IJ.O fl.lJOE+O{J (I.OOE+OO 2A6E+DO 4.0 n.-001;:+C!O O.OOE+CO O.OOE+GO O.OOE+CD >D.OOE+C(l ::LD2,E+OD O.OOE+OD O.OOE+OD O.GOE+OD 1.77E+OD N E 5.fl 0.,[)0E+-1'.Hl D.DOE+C~iJ HI O.OOE+CO .Cl.!OOE+fr.O O.OOE+0-0 fl.OOE+OD .£LOOE+OO 1.3.3E+OD m.o flA)OE+C*O C.DOE*O:tJ D.aOE+Ofl {LOQE+IJ,!J U.OOE+DD 1,D3E+OO Assessment 0,1!;,i Re~.dlts. Sa,1~ w Ft;,. M F Grand Gulf 10Miles Mi:mitcred R1;"l1;-:;>is!;' C..qQ5JCl17 144124.IJRll K H J
No PAGs Exceeded. Iodine- 1-72E-CJ2 (12.8%) Revievved B y : ~ - - - - - - - - - - - - - - - - - - - Noble Gas. 1. l6E-0 t i.86.E,%) G:a:i~ *Stuff i' 2.0.1.0 EP-CALC-GGNS-1701 Page 57 of 71 Revision O
Attachment 4 AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent AXM 3 - General Emergency Dose Assessment Grand Gutt Method: :[)etailed As51"smiait-~Monvtor-ed R~ea5ie Release P~thway:: <RCS~.,. ~'TS Bldi? <HVAC Fll~l;!I;'~> ~En..,.~;; - PRF; S.OO:E..02 Co,11i'iain~t HiJt - -Ni'A Qi;int,1i11ine11t SpRI~: _" t::Jl.A Sww PQQ] $t;;:lbtJli; ;; NiA S'llfery fi!~i::;:. NlA f:,VA(: Ffi:1iera: ;;; W~~ kl'x. 9ldg HUT:: .;;: NJA Turtiine 61(19 HUT:;;;- <2 P~~ RadW.;;1,.~ti; 'l;lkf,g H.IJT:; ;:; NIA Source Tem11.:- R~1.'!ei:er Cor~ Aideri11 ~ C~d OnSite- lti!flli:'-r Time Aft~r $.tO fhh:mm}: 1;00 l.ll.Cind: From 2.70.;. @ 4.4 n,~;h Release Ovrati>Ql'I- {hh:w.m); 1;(10 ETE{hl'l:mm);. [N,t,A J S~bnity Class: D P'r~lpita~it:m: ~:(:l!{;!e-
-o.7 1.3'1E+02 2,fi;;E+03 HI 1.2{15+01 1,85E-t01 7.44E+01 f.2lE+03 L5 1:12e+o-0 ,4.~4e:*oo. 2.5tE+D1 fU30E+Q,Ir 6.32E+02 p 2.0 !L1i!lE+OO UCIE;+OO 1.57E+01 2A19_E+01 3Jl4E+Q2 3.0 3. g51;-+0'0 ~fll!E~-0 R15E+[MI 3.62E+Cl0* t60E+01 2A6E+02 4.0 :a.sie:+oo 2.,,e~oo fJJ:h5E+11-0. 2.94E+OO 1..3tE+01 2J}3E+02 5.(l 2.a~e+o'1l 1;1~E+OO 7;D3E+OO 2.!53E+OO 1.14E+Q1 1J7E+02 TO HllE+IKl 1.2ee~..o 5.30E+o0 1.86E+OO 8A!E+OO 1s33E+02 10_0 a.r.ge:.01 4. tOE+O{I 1.39E+O.O fl.3-7E+OO 1:03E+02 M
Reviewed Sy: P~e 1 of 3
- EP-CALC-GGNS-1701 Page 58 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent AXM 3 - Site Area Emergency Grand Gutt Method: l).eeail~d Ass,essme,nt
- Monlti:m,ci Rell!:~~
Release P3.i1W,';ly' ><:;RC$~"" "':To Elklg> ,.r;f,,f,j/4,C Fil~er.S:,* '<.:Eriv> ?RF: .S.OOE-02 Condainm*;'flt HVT:;:;; N.'A Cont.:3inm,ent, Spr;,11~-;.: == NiA $',rpp Pi:;r.:;,l $t.,1t*,ll~: =.NJ..\ $3fet;, Filters; =NII, H!JAC F~1trs-: ;:.; 'No\'k:t!IJ /<.\.Jx Bldg HUT: ... ~Ji.A Turbine Elldg HUT; =.::::;: l'-J{.ru;:rs RadW~t!? E!M9 Hur : :. N,',A,
- Sciurce Te-mi: Re-~cr,cir Car;;, ,.:..f.~ici,;;,,n{ ~ C!:s.d OnSice lcw!l?r Time After $.!CJ (hh: mrn}.; 1:00 1/vind: Fn:;1rf1 :m::,,;,*,sg, 4.4 rnp:;h ETE (hh:mm): (Ni.A. J Stabi iity C!...1s$: D Pre.:i pir.a!ion: Nor.E Monitor. AXM oh :3 Disfan-c,;; E:q:i,:w1..1se, Eini;;:rh~1 ,*.. *,c
- '.: .. TEDE ' ' CDE Ft:.!e: f:u(t~ ... DD$ Gmur,d: DD l;: Thyn:,ja (Mile::;) \n11,fVr,r, .(f'.li\R:tfEX,/_ , rniR-1:!ni;, t;mR:eni) {rnR4n'!f) trnRE:111) A 3:8. :J,lSE;:+-DD_ ., "I.Ar: ..:..f"lI
-.*.- **- .... """!. 1.98E+tJ1 '.(.70E+DiD . 2.6f7E+01_ _5,G2E:rC\? .. R B 0.5 J,Ot.E+C{l !;!.l:£::,i,,0-Q 1.*UE+IJ1 5.4.SE+OD .2.14E+01 3.. 58E+02 0.7 .2.0,;:;f:::+,(,l'Q 1.2~E+O:i'J B.5l5E+Ofl 3.3tE+OD t311E+01 2.. 15E+02 i.O 1.WE+£Hl 1 !tl:!E.-0 z 4.BOE+D{I 1.BE,E+UD 7.44E+OD 1.2IE+02 1.5 1:i2e..1H ~-"1415:.1:}i 2.::HE+Oil Q.)31JE-D, 3.112E+Ob :a:::1.2E+o1 p D 2.{l 5.1.f.JE:.tl~ 33-ilE*!'.H 1.57E+DO BJ}lE~(H 2.49E+DD 2L94E+01 3.{l j,(f5E;(H -~)l3EAH 9'.7>f.iE-Dt 3.. 62E-1Ji 1.60E+OO 2Ai3E+01 4.0 3.l lE.Ot 2'.1HE:.tlf B.G5E-01 2.£14E-Q; 1.:HE+OCI .2.03E+01 5.(1 2.83E~*Oi L7i;!E,Om 7.D3E-D1 2.53E-CH *1.14E+CiD 1.77E+01 N E 7.(1 2.C1E,(H L:26EA); 5.3flE-n; 1.8::iE-DI 8A1E-O.t 1.3;3E+01
'll[}.O 1.3-56:sl.lffi (l.{ICiE*0:0 4.10E-Dt l.3:9E-m 5.4'9E-D';J 1.03E+01 Assessment 0..10 R~~lS. $,lM'li'cl w ~~. M F Grand Gulf 1OMJ!;_;.5 Mr.mitc-f,;;cl R12;!>i':?1S-: ~i)e,;;Q 17 D801'f,B.UR11 K H J
.. Np.PAGs- &ceeded Partii:::,Jat-e lodfne- 1. 72E-O i (12.8%)
Reviewect S y : - ~ - - - - - - - - - - - - - - - Noble-Gas L16E+OO (85.5%1 G.:arn:! *8!Jtr ;' :2.0.1.0 EP-CALC-GGNS-1701 Page 59 of 71 Revision O AA1.1, AS1.1-and AG1.1 URI Calculations Turbine Building Vent AXM 3 - Alert Dose Assessment Grand Gutt Method: Detailed Assessme-nt
- Monitored ReJ~ilS<I?
Release Pi;t~hway: <RCS::,. 48 .6klg:> ..r;HVAC Fil~ers:> .,:EM~ PRF: S.OOE~0:2 Co.11,rainm<;>nt HVT~ .i N.'A Cont:fin1Y1Jl;lnf Spr;11~: .. NlA Sii:.i>PF~ St:.1t-.as: :;: N.tA $;;:sfel)' Fi!tJ;ri: :;::: NlA H'JAC Ff!l~~:;:; Wooo.r.g Aµ)( Bldg HUT:.;;: N)A Turbine Bldg HUT:;;: <2 !'-l,tiu43. RadW11.t1~ BMg HUT:.;:; NIA Source Te(l't\: R~~etor Corr:; ~,;1'\'t - Ciaid OnBite- lo\*,oer Time After S.'O (hh:mm}: 1::00 1
/lffnd: Frc,ri'l 27,Q~ @4.4 m~*h R-elease :O\..Ar.MJQl'I, (hh:1'11mt 1:0C ETE {hh:mrri); [N.'A] $.Ubl lity C!,ls!;i: D Preci.pita,ron: Ncme Monitor: AXM ,ct, 3 Readings; 1.90E+01.cpm Distin~ . Eiposur,i' a1ii':'.mil.. -lriha!aii,::.n Q:-pcsi~il::*n*. CDE>*
'. . .* Rate iF.(uft1,;- . t~ EDE G;u1.m*:fJJD E i:ti~r6icj' *Circl>::" .:i\stJl'lO,Hi are, 2; :5 ,a,!Wi 10 n-it.,.:s.
' .. *;)Miles):;*.*** A (()1,R/r1r\ : '(~iA:eiiro;. .i;'(J7J;1Rii:!m} ' *~n1R:en7,)' .* ,<rnR.;fi7,) A R B 0.5 3JJ,5EHH t.MEAH 1A 1E+Oll BABE-Di :2.14E+C(I Zt58E+01 1.3\E+OO 2,15E+0*1 i.O L29E,0! *tLOOE+OO .4.8{1E-CH 1.f!5E-O~ fi.e,5E-CH 1.2!E+01 251E-IJ .... H'*..l'fi.C Fikees> .:::En..,::. PRF: 8.00E-02
():,mainm~t Ht.FL ;:: N.'A Cc,ntainw~nr, $prn~'$: =rJfA Supp P~*l $ta~.i;s: :;:: N1A Safety Filters.: -= Nil, H*,l,.!l,C Fii,1..,rs.;::;; Wor~.-~,,g /;J.1x Bldg HUT: -:;-. r,,trA Turbine BldfJ HUT: = < 2 Hcu-15 Rad'."-,'2;~,te Bta9 HUT: =J.J.'A Scurce Te-n*ti: Re.'lc;tar r.;::or£;-Ae.ckl~, - Ct3Jd On.Site- t.c,wo;:,r Timec After S.*Q (hh:.mrn}: U}O *Nind; Fn:,m 27D* *@ 4.4 nipil Release- [h.,,1r..iiti,i:.m (hh:1"1-.mJ'. 1:DO ETE {hh:mrn): (N/A J Sta bi litf CL,ss: D Preci pi~.:.iion: t*k,ne Ri::adings: 1. 13, i E+OB cpn1i :Fkrwrate: 5DcflD CFM Dist::m*~ EzpC--5<.;;f,?; E,.,-:,e.rn.A; 1r,n.3:).cd,::.n Dep0si~ion TEDE r;:DE .. Ev~cuafo:in. .Areas Fra.m*o ta 10 Miles R.:,~e P :ur.1"" DDE CED':: Gn;:.ur::j DD E T_h*1*roid
,:Miles;, ~rk~.Rmn (i,;i'.R:1:~1'1 <mRem:, ,;m.Ren*,) ~rnR:eru) (rnREn1) A
~;!:i;O fE.f._Ol R B S.B; ::LnE+Ci1 _2:!3E,..f!i 1.1lSE+D2 7.70E+D_1 2.Q7E+02 0.5 3.*04E+G1 1.S4Et-Ol 1AtE+0'2 :,.48E+0.1 2.14E+02 3.57E+03 0.7 2.-0*I:;E+(}l 1,2e.E+Gq 8.52E+01 3.30E+G1 1.31E+02 2.1t::iE+03 Hl 1.2SE+(]l1 7.9~,E+OO 4.80E+0,1 1.85E+Dl 7.44E+01 1.2~E+03
,,5 7AJ.eE+!'.Jb- 4i.44E-i,Q;Q 2.51E+fr1 9.cJOE+OD 3.92E+01 6.32E+02 p 2.0 5 .. 1136+00 :s::re;t2+i::m 1.5oE+Oo1 5.ME+GD 2Al8E+Ol 3.Q4E+02 3.f! '3.93E+CO 1.e2e+oo :9.76E+JJD 3.62E+OD 1.60E+01 2.4i3E+02 4.0 .3.30E+l'.l0 2-_1,E-rGD 8.05E+Ofl 2.ME+OD 1.31E+01 2.G3E+02 2 ..$2E+O{l 1 7'~E+fr0 2.53E+D-O 1.13E+Ol 1.77E+02 N E
- 5.0 7.0<3E+OD 7.fJ 2:00E+OO i .1,!:'.1E+l'.lll 5.29E+flfl 1.86E+O:D .8.40E+OD 1.32,E+02
*rn.o 1.35E:+CO fL7o~-tJ.; 4.10E+OD 1.39E+O{I 6.37E+OO 1.D3E+02 As-s.essment Dqu, RH.J:1s. $3~\i:ci ,.o Ft.~. M Grand Gulf 10Mi1;s Maf!it<<ect R~lE-JrSi:; u:4051017 141715.IJRI?
K H J PAGs Expeedecl in Designated Areas lodfne- 1.72E+OD (1.2.9%} 1.15E+01 (88.4~,'ii) Review*ect B y : - - - - - - - - ~ - - - - - - - - - - - - P,:;J.; i of 3 Gra'.l~ *:?u!f ,* 2.0.1.0 EP-CALC-GGNS-1701 Page 61 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent AXM 4 - Site Area Emergency Dose Assessment Gr~nd Guif Mettiod: Det.1ile,d Assecssme-.nt - Mcmr:tored Rel,i;-~s,e ReleaseP:;1.~liway '<RCS> *46 . 6kl~ .:;HVAC Fihers> <Env~~ ?RF: S.OC!E*O::? Co:11rainr11.(';n~ Ht.fr; .i N,'A CQnt.;Jinrr.ent Spr;it~: .:; NtA $,,pp p~ $t;Jt*41i;; :::: NJA. Safe,:Y Filteri; *.. N!A
. HtJAC Fii:i~;;; Wim."~ A!Jx Bldg HUT: ;:: NfA Tu~inl';' Sldg HUT:* <;2 r-',ouiS. R.;Jd'iN~$b:: &Jc!,g HUT'. ;::; WA Source Term: Re-:;,ctar Cori;; Ae,,ciai,;-rn - Ct.3d OnSlteto,,,,,er Time After .5,10 ihh:rnm}: 1:CO lflnd: Fr<;:,m .270'1<@ 4.4 mp::,h 1
Release- [hJration {hh:m,m}; t{JO ETE. {hh;mm): (NIA). S::tabl lity Class; 0 Precipita,ion: NClf"o~- Monitor: .AXM (;t, .; Fl:owrate: 5DOO CFM TED.E: *. (;OE;-_ Grour:1:fDDE Th:;:rqJa :
- Circle &$,nres are 2. ti a,n,j 10 tnl1?s.
.. lnMv.~n,;;) . {n)8i~ni). . 'f.foR~fiH *. A R B SJ~. }., 7.0E:-t;Ofl, ~~ . ,:;\9?E-:+:0.1 0.5 1.ME+OO 1.4SE+Il1 5.48E+fJ{l 2.14E+01 3.57E+02 0.7 2.nee+on 1,2e;E:+CNl 8.62E+Ofl 3,30E+G{l 1.3!E+Oi 2,16E+02 rn 1.28E+Cn 7J;}fJE-Oa 4.BOE+IJ:O 1, 85E+O{l 0
7.44E+OD 1.2:tE+02 1.5 ;;oae,.iJt ~A4E*O! 2.5tE+OO Q..dOE-D; 3.Q<2E+C*b fL32E+01 p D
.2.0 6.165-CH :l.:lt!E*lln 1.515E+OO 5J}1E-D1 2Al8E+OO 3.94E+01 3.0 fG3E-(H l-6~E-OV . 0.76E-Dl 3.62E-01 *'1.60E+OD 2A5E+Q1 4_.'[l -3.30EHit .Z" 1 lEH)t 8.05E-Dt 2.ME-D; 1. 31.E+IXI. 2.D.3E+01 N E 5.0 :ui2e:.om L79E*CH HJ;3E-DS 2.. 53E-Ul 1.13E+OO 1~77E+01 7.(1 2.ooe.-m t.25E*Off 529E-D1 *t.8.6E-01 8.40E-D1 L32,E+01 mo 1:3.5g.o, OAlOE+OO 4.'10E-CH t.3fJE-Di !;j,.4'9E-01 1.0JE+01 Assessment 0Jt:Ji R,ss:J~ S;:;,1 iO F&;-: M F Grand Gulf 1OMile~ Monit<<ed' Rel!S'~~ 0405:2017 14 lBCiO.URl7 K H J
R9'2E"-03 (D.7%} lodin~ 1.72Ec.flt (12.9%) Noble Gas U5E+OO (86.4%}
.Reviewed B y ~ - - - - - - - - - - - ~ - - - - -
Graml Gulf.I :2.0.1.0 EP-CALC-GGNS-1701 Page 62 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Turbine Building Vent AXM 4 - Alert Dose Asse.ssment Grand Gulf Wedn4;'~da;y, April~. 2017 14:Hl Method: Uetafle-d As.!>e,s.sm1:n.t - Monrtor-ed R~e~s,e R,elease P~~h'i.'1.'J'.)". <:RCS:;;. .;:1fE, Stas> '",H'*l..:..c Fil'.i;:l."s.?* *<Erw> PRF: 8,0GE-02 v::,mainri\<;'flt Ht.tr:. = N,'A ,Cp11tain1"l'.,;1nt Spr;;t~'$: =NIP, $-u:pp PD'.$ $1:at*,r,s: :::; ~M S;,1fel:)' Filters; ~ NY*. H*.1.1!..C Fiilo:rs:::: Wark.n,;;i Am: Bldg HUT: NIA = TurbinE< Slag HUT:= <:2 r'~u~ R.:id'N::,r,te 8-Ed-g HI..JT' :::; !-.J,'A
- 3ouri::e Ts-mi: Ro:.:?tctcr Cori;- A¢,.,~l.e-m - (;Lid On3i:e- Lower Tr me, Ar.er ::;,ip fhh: mrn }: l'C-0 Wlnd: From 27[)~ ,@4.4 n,~;h R~le:ase Dvr.:.ti,011 ~hh:n1m): l;OO ETE (hh:mm): (N..'A) Stabi !ity C!aJtS!:,: 0 Preci piea,ion:: t*Jar,e Monitor: AKM ch 4 Ri:iadings:. 1. IJ.1i E+D4 ,opn1, Rmwate: 50.{)[} CfM Dist3n*.:>= Expoo,.*:fr.: E~::"e-!1~1:s.t 1r;i),:;Ja~ion Def::-0si~ion TEDE. CDE.
- R~1~1:, DOE CED:E (:imur:,a',i:) ::: E Tr1*>'roict Circle, &st.'inc.~s are 1:. 5 -31'1*:+/-: 10 m,~,;,s.
,;MHes) ,,:mPVr,r:< f<:1,*R:.;.i:r) (rriRi:m 1 ,:111,R£<f!1') \mRien)) {mRi::m) A S.B. 3.i7~-I} 11~E*,e,. 1.9':;E+OO 7.rnE-CH 2.97E+OO R B 0.5 ::-04E-CH iB4E-C,t 1.4\E+O{l 5ABE-G, 2.14E+OD 3.57E+01 0.7 ::fO,EJE-Cl*, L2e!E-!'.}~ 8-52E-Gt 3_:3{1E-Ol 1.3 ~E+C(l Hl L2BE*O*, 0:0('.IE+{l.Q 4.BDE-D: 1.8.5E-O*:i OB5E-Dl 1-2:E+01
~.5 {J.,cim~+ci*o :0.tlbE-i.i:i::ri 2..fr1E-Gll .fLOOE+QD 2.f.:,1E-Dt 6.32E+OO p 2.0 o.noE+cm tl.OC1E:ti:l'l'.l 1.5*5E-Dt
- O.OOE+D{I l.5flE-llli 3Ji4E+OD 3.'D *:n.oOE+Od O:bOE+M. fl.lJOE+OD O.OOE+O-D *D.OOE+OO 2Ai3E+OD 4.0 fl.tlOE1--C:*(l O.fl,:iE+CO fl.OOE+ClO {l.OOE+CD *D.OOE+C(J 2.D3E+OD 5.0 {LOOE+O*O D.OOE..-fliJ {J.DOE+0-0 r.J.OOE+Ofl fJ.OOE+CiD 1.77E+OD N E 7.0 O.OOE+{JO *O;OOE.*1,'JO O.OOE+OD O.IJOE+O:[t O.IJOE+OO 1.3:3E+OO m.o OJli:d:;+QO ;CUJOE+Q. !J, {UJOE+G{! O.OOE+OD O.IJOE+OO 1JJ3E+OD Assessment 0,1.tl Re~1r. $.::,u,e,ci 'It) Fs;-. M Gr.3rtd Gulf 1DMil..-s Monite,re,a Rel'i"-3~ 040.':i:";017 141852.URli K H J
No PAGs Exceeded Partictiat-e K9'2E-£J4 (0.7%} Iodine- 1.72:E-{12 (12.G%) RevielNed B y ~ - - - - - - - - - - - - - - - - - - - - - Noble- *3as 1.15E-m ;,86.4%) Gra.1:1 *:?i.Jtr i' 2.0.1.0 EP-CALC-GGNS-1701 Page 63 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event'via Aux*Building Vent SPING 7 - General Emergency Dose Assessment Grand Gulf Taesday,February 6;2018 21:12 Method: Detailed Assessment - Monitored Release Rele_ase Pathway: "".SF> <Under Water>*<AUX Bldg>.< HVACfllters> <Env-... PRF: 8-00E-04 ContainmentHUT: =NIA Containment S-pr.ays: =NI~. Supp 'Pool Status: =NIA Safety Filt-ers:. =NJA HVAC Fil'.ers: =lflorkin,g Aux Bldg HUT: = -e: 2 Hours TurbJn,e Bldg HUT:*= N.1A RaclWaste Bld,g HUT: =N.'A Source Term: Spent fuel Accident - Under Water *oamage: 0250 % OnSite lo\*ter Time Since Irradiated (hh:rnm):.80:00 Wind: From 27D" @4.4 mph Release Duration (hh:mm): 1:OQ ETE (hh:rnm ): *[NlA ] :StaiJiljry Ciass: D Predpi_tatron: None Monitor: SPING ch 7 Readings: 6.43E+06 cpm F!owrate: 24720 CFM
*Distance
- Exposure. ?fa:tern.:i1*. ,_ 111fk11ation QepO$ition, ~!;!2E *'cPE..
ij1te ,:PlJnie ODE, ;CEDE: Grm:i~'d'b°DE
- JIW.i:o_i~* .. Circle dLstanoes are 2. ,5 and ta mife~.
':(r.hRinrk :* :(r,i'R.'en-1} ,*:.(ntRen)), * ,Jn1ij~h~) * \( nifi~:ri{), -*
- tn,~e~{:,
- R A
8
*.,.~J:]OE:+Jl,2 J.90E+0*1 ~:BB[;:!-_D_O_, .:,:2:t"o91;:*.:i:0'3;;..".'. , __4)~E.=tJl2 0.5 1.13E+D3 7:92E+02 1.38E+01 4.27E+OD 8.1GE+02 3.07E+02 0] 7.40E+D2 5.20E+02 B.48E+OO 2.62E+OO 5.'3*1E+02 *t,8BE'+02 1.0 4.52E+D2 3:16E+D:?, 4,84E+DO 1.49E+OO 3.2?.E-t-02 1,0BE+02 1.5 2.44E+02 1.71E+02 2.56E+OO 7.93E-01 *t.74E+O;:! 5.68E+01 p 2.0 1.98E+D2 1.39E+02 1.60E+OO 4.'94£-01 1.4:lE+02 3.55E+01 3.0 1.39E+02 9.56E+01 . 9'.22E-01 2.76E-01 :9J58E+01 2.06E+01 4.0 1.HE+D2 a~nE+o1 7.72E-D1 2.30E.:.01 8.23E+O*l 1.72E+01 7.18E+01 tA9E+0*1 N E 5.0 1.05E+D2 7.09E+01 6.7DE--01 1.97E-01 1~0 8.:36E+01 S,83E+01 5.47E-01 1.59E..:01 5.90E+0*1 1:22E+01 10.0 6.44E+D1 4.41E+0*1 4.09E-01 U7E..:01 4.46E+01 9.12E+OO
*k:-sessment Data Resuhs Saved to File: M F
-Grand Gulf10Miles Monitored Release 020021118 2112:59.URl7 K H
.J Revievred By:
Gr"1lc,Q! Gui! I J;Q, '.0 EP-CALC-GGNS-1701 Page 64 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent SPING 7 - Site Area Emergency Dose Assessment Grand Gulf Tuesd{ly1 February G, 2013 21:15 Method: Detaifed Assessment - Monitored Release Release Pathway: <SF> <Under Water>* <AUX Bldg>< HVAC FJlters> <Env> .PRf: fLOOE-04 Containment HUT:= NJA Cont3tir:ment Sprays:= N.r..:.. Supp P-ool Status:= NlA Safe!j* Filters: = NJ'A H\IAC :riliers: =Wor.kin,g Aux Bldg r.:UT: =-< 2 Hours Turb.ini! Sieg HUT:= N/A R3.dl/faste E!ld1g HUff: =N.'..!.. Source Tem1: Spent Fuel Accidem - Under Water Dam.:lge: [L250 '}:; OnSite Lower Time Since Irradiated (hh:mm:i: 80:00 1Nind: from 27D 0 @ 4.4 mph Re!eas,e Dt1ration {hh:mm): 1 :00 ETE (f1h:mm): [f,JJA] StaEJility CfO$S.: D Precipitation: None Monitor: SPING ch 7 ReGding:;: 6.43E+05 cprn Ftowrnte: 24720 CFM Dist:lnce Exposure faterneil l11hjlaticn Deposition 1ED,E CD.E Evacuation Are:as from O to 1 Miles o Rate Plume DD=. CEDE GrorJnd ODE Th:;lroid Circle distances .are 2. 5 and HI mili's. (Miles) (mRJhr) (rnRem.) (mRem) (mRem;, (mRem) (mRe1_n) A
*J,90E+OO 5.88E~D1 1.0GE+02. 4.22E+01 R 8 S.B. 1.10E+02 9.80E+01 0.5 1."13E+02 7.82E+D1 1.38E+OO 4.27'E...cD*1 8.1GE+01 3.07E+01 0.7 7.40E+01 5.20E+01 8.48:E~D'l 2.62E-01 5.3*1E+0*1 1.B8E+0*1 1.0 4.52E+01 3:i6E+O*J 4.ME-Ci"l 1.49E-4)1 3.23E+O*J 1.0BE+01 1.5 2.44E+01 1.71E+0*1 2.56:E-D*l O.ODE+OO *U3E+01 5.6BE+OD p 0 2.0 1.98E+01 1.39E+01 'L60E-D1 O.ODE+OO 1.40E+O*J 3.55E+OD 3.0 1.39E+01 9.56E+OO O.OOE+OO O.ODE+OD 9.5-5E+OO 2.06E+OO 4.0 1.lTE+O*l 8.13E+OO D.OOE+OO O.OOE+On 8.13E+OO 1.72E+OD 1.05E+0*1 7.09E+OO O.OOE+OO O.OOE+OO 7.09E+OD 1.49E..-00 N E 5.0 7.0 8 ..36E+OO 5.83E+OO {li.OOE+OO O.OOE+OD 5.83E+OD 1.22E+OO 10.0 6.44E+OO 4.41 E+OO O.OOE+OO O.ODE+OO 4.41E+OO 9. t2E-01 Assessment Data ResuJ15 Saved to Hie: M F Grand Gulf 10Miles Manl\ore-d Re!-e2ise {l2G5W18 21 rn27JJRl7 K H J
NI) PAGs Exceeded
---- ~-~
I. ~'fa,.. " .:; "' .... * ~ " """' ~ _;+' * ;( , Release Rates*(Ci i sec) Particulnie i .43E-03 (0.0%) IOdine 3,4 iE~D3 (0.0'%) Revie*wed By: ____________________ Nonie Gn!l Gr;;.n'<(li Gui(! .2.0. ! .0 EP-CALC-GGNS-1701 Page 65 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent SPING 7 - Alert Dose Assessment
~rand Gulf Tuesday; February 6. 2018 21:16 Method: Detailed Assessment - Monitored R~{ease Release Pathway: <SF> <Under Water> <AUX Bldg> < HVAC Filters> <Em*> PRF: B.OOE-04
_Containment HIJT:.= 1"'1A Con:tainm~nt Sprays:= WA Supp Foo! Status:= N/A Safety Filti?rs: = NJA
,HVAC_ Filters: = Viorkin,g Aux Bldg HUT: =* <.:2 Hours Turblne Bldg HUT: =NfA Rad'l1faste Bldg HIJT: *= NIA Source Tem1:Spent Fuef Accident - Under Water Damage: 0250 % OnSite Lower Time Since Irradiated (hh:rnm): 80:00 lNind: from 27D @4.4 mph 0
Release Duration {hh:mm): 1 :DO ETE f.hh:rnm): [NIA] Stability Class: D Precipi_taJE.on: None Monitor: SPING ch 7 Readlngs: 6.44E+G4 cpm Ffowrnte: 24720.CFM
*,Ojstilnce -:* ,ExposLJre Exter11al ..* *,*1hh~11atio n .Deposition '.
JfP~ CQE - '. Plun,e ,_.ODE*:
.: c-:t]~?te ;:\, :'.,:/";*_;* ;\<'
,,,'c:'.cC:;EDE *, Gtritind:DDE
~1~[Ji5 . *. . ,t~yr~W\:. Cirele-. di.stance,s are 2. 5 an.d iQ mil-es.
;~:r\~.~fleif ' . , }(mR/"h[) , --,!(roRemJ :(fojijen;1): .Jri11:{eih)* :.* '(ii~f:(.erm....
R A 8
... s,s. ~ ~-- 1,4QE70J_ ,,_?'J\lE;--:00, .. .JJ*_Q~~:01 ., .JU!()E~@-. ,., 1_,QJJ:E:i:O.L ,_4:,2:JJ~~!]D 05 1.13E+01 7:92E+DO 1.38E-01 0.00E+OO 13':0SE+OO 3.0BE+OO 0] 7.44E+OO 5.20E+OO O.OOE+OO O.OOE+OD. 520E+OO 1.89E:+-OD 1.o 4.52E+OO 3.HE+DO O,OOE+OO 0.0-0E+OO 3.17E+OO 1:0BE+OO 1.5 2.45E+OO 1.71E+OO O.DOE+OO O.OOE+OO U1E+OO 5.72E~D1 p 0 2.0 1.99E+OO 1.39E+DO O.DOE+DO {toDE+OO 1.3&E+OO 3.56E-01 3.0 1.39E+OO 9'.6DE-01 O.DOE+OO O.OOE+OO 9.6DE~01 2.06E-01 4.0 1.18E+OO B>l7E-01 O.DOE+DO O.OOE+OO 8."l?E-01 1.73E-0'1 1.05E+OO 7:1DE-01 o~ooE+oa OJJOE+OO 7.10E-01 150E.:01 N E 5.0 1:0 8.40E-01 5.83E-01 O.OOE+OO D.OOE+GO 5.83£-01 1.22E.:01 10.D 6.4BE-01 4.4*1E--01 O.OOE+OO O.OOE+OO 4.4*1E-01 D,OOE+OO
- Assessment Data Resul:s Saved to File: M F Grand Gu1f10r<i1iles Monitored ReJ.ease 02CJ.fl2D'18 21 t639.IJRl7 K H J
3/fiE-04(0.0%) Reviewed By: il64 +01 100.0%) EP-CALC-GGNS-1701 Page 66 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent AXM 3 - General Emergency Dose Assessment Grand Gulf Tuesday~ February 6, 2018 21:19 Method: Detailed Assessment - Monitored Release Release Pathway: <SF> <Under 'Water:> <AUX Bldg> < H\IAC flltern> <EnV.> PRF: 6.00E-04 Containm~nt HUT: = NJA Cont.:iinrn.er.t Sprays:= Nt.G. Supp Poof SJtatus: = NiA Saf-er.y Filters:= WA H\J AC Filters: =1Norkin,g Aux Bldg HUT:= -0:.2 r.ours Turbkte Blc'g. HUT: =N.i'A Rad!Naste Bldg HIJT: = NI.O. Source Tem1: Spent fuel Accident - Under \Nater Damage: D.250 % OnSite Lower Time Since Irradiated (hh:rnm): 80:00 Wind: from 27D 0 @ 4.4 mph Release Duration {hll:rnm): 1 :00 ETE (ilh:mmj: [NU\] Stal)ilify Cfce;s: D Predpitatron: None Monitor: AX M er, 3 Readings: 4.3BE+05 cprn Flowrate: 24720 CFM Dist-wee Exposure E:~terncil I11h.:; l,3tio n Deposition 1.EOE CDJ: Evacuation AreM from_O to 10 Miles Rate Plume DOE CEDE Ground ODE Thyroid Circle diMances ar.e 2. 5 and iii) miles. (Mtles) * (mRthr) *.(mRem} (m.Ren*i; (fnRem;, (m!Rem) (tnRerr,) A R 8
*s:a. 1.40E+03 9.80E+02 .. "L_90E+0*1 5.8BE+OO :*t ,OQ~+O 3::. 4 .. 23E_+02 0.5 1.13E+03 7.92E+02 1.38E+01 4.27E+OD 8.1C.E+02 3.0BE+02 0.7 7.40E+02 5.20E+02 8.48E+OO 2.62E+OO S.3-:1E+02 LSftE+02 1.0 4.52E+D2 3.17E+02 4.84E+OO 1.SDE+OO 3.23E+02 1.08E+02 1.5 2.44E+02 1.71E+D2 2.56E+OO 7 .93E-01 *USE+02 5.72E+0*1 p D 2.0 1.98E+02 1.39E+D2 1.60E-:-OO 4.94£-01 *1.41E+02 3.56E+01 3.0 1.39E+D2 8.57E+0*1 9.27E-01 2.76E-01 9.70E+0*1 2.06E+01 4.0 1.17E+02 3.15E+01 7.75E-13'l 2.30E-ui-1 .S.25E+0*1 1.73E+01 6.72E.~0*1 1.'97E-01 7.18E+01 1.50E+0*1 N E 5.0 1.05E+02 7.09E+01 7.0 8.40E+01 5.83E+01 5.47E..;[)1 1.59E-D1 5.9DE+01 1.22E+01 lD.O 6.48E+01 4A1E+0*1 4.1DE-{H *U7E-t:n 4.46E+01 9.14E+Ou A.!:.sessment Data Resul-.s Saved to File: M F Grand Gulf 10Miles Monir.ored Rel':ase Q2[}fl2G18 21 i91RURl7 K H J
PAGs Exceeded in Des.ignat:e-d Areas Release Rates-(C:i J sec) P::Htieufate i .43E-02 (0.0%J Iodine 3AH:.-02 (0.0%) Reviewed By: Pa;e 1 of 3 Gr:.:m,tt Gui( I J.O. I ,0 EP-CALC-GGNS-1701 Page 67 of 71 Revision O AA 1.1, AS1 .1 and AG1 .1 URI Calculations Fuel Handling Event via Aux Building Vent AXM 3 - Site Area Emergency Dose Assessment
- Grand Gulf Tuesday~ February 6;2018 21:20 Method: Detaired Assessment - Monitored R.elease Release Pathway: <SF> <Under Water:> <:AUX Bldg>< HVAC. Filters.> <Env> PRF: KOOE-04
*containment HUT:= WA Con1:ainment Spir.ays: = N.'A SUPP Poot St3.tus: = NIA Safet;* Filters:= NlA HVAC Filters: =Workin~ Aux Bldg HUT: = < 2 Hours =
Turbin,e Bldg HU"f: N!A RadWaste .Bldg HUT:= NIA Source Tem1:.Spent Fuel Accident - Under '!Nate( Damage: {t25D % OnSite Lower Time Since Irradiated {hh:mm): 80:00 Wind: from 27D"@4A mph Release Duration (hh:n1m}: 1:00 . ETE {hh:mm): [N/A] Sta~ilify Class: D Precipitation: None Monitor:*AXM ch 3 Readings: 4.3BE+04 cpm flowmte: 24720 CFM
.cc;t*
J:ltYf9.f.: <~-,
;,*t1hR~iii' Circle di-stances are 2. 5 and 10 mil.es.
A R 8 0.5 1.13E+D2 7:92E+01 1.38E+OQ 4.27E-D1 8.1DE+01 3.0BE+01 0.7 7.40E+D1 5.20E+01 8.48E~01 2.62E-0*1 5.3tE+01 *t.89E+01 1°.D 4.52E+D1 3.17E+01 4.ME-01 1.SDE-01 323E+O.*t *LOBE+0:1 1.5 2.44E+01 1.71E+D1 2.56£-0*1 {HIOE+OO t.74.E+01 532fa00 p 2.0 1.98E+D1 1.39E+D1 1.6DE-01 {tGOE+OO 1.40E+O-t 3.56E+QO 3.D 1.39E+01 9.57E+OO o~ooE+oo O.ODE+OO 9.57E+OO 2.06E+OO 4.0 1.17E+Of 8.15E+OO O.DOE+OO O.OOE+OO K15E+OO 1.73E+OO 7.D9E+OO O,OOE+OQ O.OOE+OO 7.09E+QO 1.50Ei-OO N E 5.0 1.05E+01 7.0 8.40E+OO -s.83E+OO O.OOE+OO O,OOE+GO 5.83E+OD 1.22E'+00 10.D 6.48E+DO 4A1E+DO O.DOE+OO O.OUE+OO 4.41 E+OO 9.14E.:.01 Assessment Data Results Saved to File: M F Grand Gulf10Miles Monitored R-el-ease020f!21l18 212035.UR17 K H J Reviewed By. EP-CALC-GGNS-1701 Page 68 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent AXM 3 - Alert Dose Assessment Grand Gulf - Tuesday, February 6, 2018 21:21 Method: Detailed Assessment - Monitored Re[ease Release Pathway: <SF> <Under Water:" <AUX Bldg>< HVA.C Filters> <Emr> PRF: 6-00E-04 Containment HllT: =NJA Con.tairnm,ent Spr.;;ys: = NlP. S,..ipp Pool Status:= N!'.A, Safe:;y Filters: = NJA HVAC FiltErs: = lJ\'orkin9 Au:.: Bldg HUT: = < 2 Hours Turbine Bldg HUT:= NlA Radlhf:,,ts!e B!d9Hl}T:*= NIA Source Tem1: Spent fuel Accideni: - Under Water Damage: D.250 % OnSite Lower Time Since Irradiated (hh:rnm): 80:00 Wind: From 27D~ ,@ 4.4 mph Release Duration (hh:rnm): 1:00 ETE (ilh:mmj: [NlA] StalJility Crass: D Precipitation: None Monitor: AXM cl13 Relldings: 4.39E+03 cprn Flowrate: 24720 CFM Distance Exposure E:~tem31 *111h,21.lation Dep,Jsition iEDE CDE Evacuation Are.as *from O.fr, 10 Miles Rate Plume.DOE. CEDE Ground ODE Thyroi~ Circle dis.tances are 2 ..5 and tlJ miles. (Miles) (mR.'l1r) *a(mRem} (niRem) (inRem) (mRem) (mRetri) A 1.onE:1:0_*1 4.23E_+OO R 8 SJ3. *1.40E+01 9_.S*J E+OO 1:90E-:(t1 O.OOE+QD 0.5 1.13E+01 7.92E+OO *l .38E-fr1 O.OOE+OO e.oe.E+ao 3.08E+OO 0.7 7.44E+-OO 5.20E+OO O.OOE+OO D.OOE+OD 5.2GE+OO L89E+OO "LO 4.52E+DO 3.17E+OO O.DOE+DO D.OOE+OO .3.17E+0.I) 1.0BE+OO 1.5 2.45E+OO *1.72E+OO D.OOE+OO O.ODE+OO L12E+OO 5.72E-01 p 0 2.0 *1.99E+DO 1.39E+OO O.OOE+OO D.ODE+OO *L3ftE+O-D 3.56E-01 3.0 *1.40E+OO 9.cOE-01 {tDOE+OO O.OOE+OO 9.0E-01 2.06E-01 4.0 1.1SE+OO S.*t7E-D1 ftOOE+OO D.OOE+OO 8:17E-01 U3E-0*1 7.10E-01 1.SOE-01 -N E 5.0 1.05E+OO 7.*tOE-01 O.OOE+OQ O.ODE+OD 7.0 8.40E-01 5.83E-01 O.OOE+OO D.ODE+OO 5.83£-01 1.22E-01 10.0 6.4BE-01 4.42E-01 O.OOE+OO D.OOE+OO 4.42E-01 [LOOE+OO Assessment Data ResullS Saved to File: M F Grand Gulf 10Miles Monitored Rel':ase 02[}621}18 212146.URl7 K H J
'*~Classification: Validate against Emergency Action Levels i * ' *Relea:seRates(Ci t see)
Pi:irticuh::ite i .44E..04 (0.0%) lodirie 3,42E-04 (0.0%) Reviewed B y : - - - - - - - - - - - - - - - - - ~ Gr.:111r.ll(fa1.1ltl 1.0.,.0 EP-CALC-GGNS-1701 Page 69 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent AXM 4 - Site Area Emergency Dose Assessment Grand Gulf Tuesday; February 6, 2018 21:24 Method: DetaHed Assessment - Monitored Release Rielease Pathway: <".Sf> <UnderWater>--<:AUX_Bldg> < HVAC Filters.> <Env> PRF: KOOE-,04 Containme-nt HUT: =l,JJA Con:tainment Sprays:-= N!A SUPP Poof Status: = N!ft. Safety-Filters:.::. NJA HVAC Filters: =Worlcinig Aux Bldg HUT: =< 2: Hours T;urbln,e Bldg HUT:= NlA RadWasti: .Bldg ft UT: =Ni A Source Temi:: Spent Fuel_Accjdent - Under Water Damage: 0250 % OnS1te lower Time Smee Irradiated (hh:mm):.eo:ao Wind: Frorin 270a *@::-u mph Release Duration {hh:nrim}: 1:mJ ETE {hh:mm): IN/A] StaiJJlily Cias:s::*D Predpitat~on: None Monitor.: AXM ch 4- Readings: 2.43E+07 epm. flowrate: 24720 CFM Circle. di.stances are i 5 ,3nd U! miles. A R 8 0.5 1,13E+02 1:38E+OO* 8.1{!tE+Q1 0-7 7.40E+D1 5_20E+01 8-48E-01 2J32Es4J-1 5.3-*1 E +O *t 1.0 452E+D1 3~16E+0*1 4.84EAH 1.49E.,.01 323E+O.*f t08E+01 1.5 2.44E+01 t.71E+01 2.56E-0.1 OJJOE+oo* U3E+01 5.ME+.00 p 2.0 f.98E+01 1.:~9E+D1 *-t.59E-O"l O.OOE+OO "1.40!=+01 3"55E+Cl0 3~0 1.39E+01 9.56E+DO O.DOE+OO O~OOE+OO .9.56E+OO 2.06E+OO 4.0 1.17E+01 a.nE+OO O;OOE+OO O.OOE+OO 8A3E+OO 132E+OO O.OOE+OO O.OOE+QO 7.09E+bo 1A9E+OO N 5~0 1.05E+0:1 7.09E+OO 7.0 8.36E+OO 5.83E+DO O.OOE+OO O.OOE+oo*. 5.83.E+OO t.22E+OO 10.0 6.44E+DO . 4.41E+DO O.OOE+OO O.OOE+OO 4.41E+OO -9 .. 12B01 Assessment Data Results Saved to File: M F Grand Gu1f1 OMiles Monitored Rel,ease 020-621118 212400,URl7 Reviewed By: EP-CALC-GGNS-1701 Page 70 of 71 Revision O AA1.1, AS1.1 and AG1.1 URI Calculations Fuel Handling Event via Aux Building Vent AXM 4 - Alert Dose Assessment Grnnd Gulf Tuesday, February 6, 2018 21:25 Method: Detailed Assessment - Monitored Rele.ase Release Patl-nvay: <SF> <Under \Nater> <AUX Bldg:,..::: HVAC Filters> <Erw:::- PRF: KOOE-04 Containment HUT: =t..JJA Con.tair.m.ent Spr.ays: =N,*,.::., Sup.p Pool Sw-,tus: = N."A Saf.er.y Fi!trs:. = NJA HVAC Fil"ers: = lNorking Aux Bldg HUT: = < 1 Hours TurbirieBldg HUT: =N/A RadW.as:te Bldg HIJT:*= N.'.O. Source Tem1: Spent fuel A~cident - Under Water Damage: D.250 % 011Site Lower Time Since Irradiated (hh:mm): 80:00 Wind: from 27D 0 @ 4.4 mph Release Duration {hh:mm): 1:DO ETE (hh:mmi: [NlA] StaEl ii ity Ckimi: D Precipitation: None Monitor: AX M cl1 4 Rei3dings: 2.44E+06 cprn flowrnte: 24720 CFM Distance Exposure E:~terneil lnh.:il.ation - Depo,:;ttion TED£ CDE Evacuation Are;as from O to 1'0 Miles Rate Plume DD:E CEDE* Ground ODE Thyroid Circle dis.t,:,,nces ar1: 2. 5 and tD mifes. (Mtles) (mRfhr) *(rnR.em) (mRernj (mRem:, (mRem) (mRem) A SJ3-. 1.40E+01 9.8*1E+OO 1.90E-{H D.ODE+OD 1,0DE+01 4.2JE+OO R 8 0.5 U3E+01 7,92E+OO *1.38E-01 O.OOE+Of} e..OeE+O!) 3.08.E+OO 07 7.44E+OO 5.20E+OO O.OOE+OO O.OOE+OO 5.2GE+OD t.Ei~E+OO 1.0 4.52E+OO 3.HE+OO O.DOE+OO O.OOE+OD 3.17E+OD 1.0BE+OO 1 _5 2.45E+DO 1.72E+DO O.OOE+OO O.OOE+OD *U2:E+OD 5.72E-01 p 2.0 1.99E+OO 1.39E+OO O.OOE+OO .O.ODE+OO *1.39E+OD 3.56E-01 3.0 1.40E+DO 9.60E-01 D.OOE+OO O.ODE+OO 9.60E-01 2.06E-01 4.0 1.18E+OO 8.*t7E-D1 O.OOE+OO O.ODE+Ofl a.-1 ?E.-01 -1.73E-0*1 5.0 1.05E+OO 7.*JOE-01 D.OOE+OO O.ODE+OD 7.10E-01 1.50E-01 N E 7.0 t:.40E-!H 5.831::.-01 ItOOE+OO D.0-DE+OO 5.B3f.-0*1 1.22E-01 10.0 6.48E-01 4.42E-01 D.OOE+OO D.OOE+OO 4.42E-01 D.OOE+OO Assessment [}ata Resuh.s Sa'ied to Fiie: M Grand Gulf 10Miles Manitore-d R.el-ease 020021)18 2.12509.U?.17 K H J
.._~~~~-*-*-~_c_B_a_s_s._if_ic_a_t.i_o_ri_:_~_'.a_l_id_a_t_e_amg_~_in_s_t_E_1_n_e_rg_~_~n_c_y_A~c-ti_o_n_L_e_v_-~_ls~'-'-'~~~~~-..I 1--~~~~~-*R_e~l_e_~_s_e._R_a_t_e_s~(_C_i_l_s_ec~]'--~~~~~-1 F'nrticutnte i .44Es04 (0.0%,)
lodlr1e 3A2E-04 * (0.0%) Revieiwed By: t, oble (3an 6.65E+Dl {100.0%} P..:1ga 1 e,f 3 Gr:ui-:i Gulf i 2.0., .0 EP-CALC-GGNS-1701 Page 71 of 71 Revision O
GNR0-2018/00048 Page 18 of 19 GNR0-2018/00048, ENCLOSURE ATTACHMENT 2 GGNS EAL BASIS DOCUMENT
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Grand Gulf Nuclear Station EAL Technical Basis Page 1 of 270
J Grand Gulf Nuclear Station EAL Basis Document Revision XXX Table of Cont~nts
1.0 INTRODUCTION
......... ;.......................... : ............ .- ...................................................... 3 2.0 DISCUSSION, ............................................................................................................ 3 2.1 Background ...................................................................................................... 3 2.2 Fission Product Barriers ................... :......... : ..................................................... 4 2.3 Fission Product Barrier Classification Criteria. ................................ .'.................. 4 2.4 EAL Organization .............................................................................................. 5
- 2.5 Te_chnical Bases Information ............................................................................ 7 2.6 Operating Mode Applicability ............................................................................ 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ............. :.................... 9 I
- 3. 1 General Considerations ...............................*.................. ; .................................. 9 3.2 Classification Methodology ...................................... *.................. :.................... 10
4.0 REFERENCES
...................................................................................... ,................. 14 4.1 Developmental. ................................*............................................................... 14 4.2
- Implementing .................................................................................................. 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ........... : ....................................... 15 5.1 Definitions (ref. 4.1.1 except as noted) ........................................................... 15 5.2 Abbreviations/Acronyms ................................ *................................................. 21 6.0 GGNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ....................................... 24 7.0 ATTACHMENTS ...................................................................................................... 27 7.1 Attach~ent 1, Emergency Action Level Technical Bases ............................... 28 Category A - Abnormal Rad Levels I Rad Effluent ...... '. ................................ 28 Category C - Cold Shutdown I Refueling System Malfunction ...................... 65 r Category E - Independent Spent Fuel Storage Installation (ISFSI) ............ 106 Category F - Fission Product Barrier Degradation ...................................... 109 Table F-1 Fission Product Barrier Threshold Matrix & Bases 113 Category H - Hazards and Other Conditions Affecting Plant Safety ........... 168 Category S - System Malfunction ............................................................... 203 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases 243 Page 2 of 270
Grand Gulf Nuclear Station EAL* Basis Document Revision XXX
1.0 INTRODUCTION
This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Grand Gulf Nuclear Station (GGNS}. It should be**used to facilitate review of the GGNS EALs and provide historical documentation for future , reference. Decision-makers responsible for implementation of 10-S-01-1, Activation of the Emergency Plan, may use this document as a technical reference in support of EAL .interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification. Because the information in a basis document can affect emergency classification decision-making *(e.g., the Emergency Director refers to .it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). , 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the GGNS Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
- Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
- Initiating conditions _and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).
- Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implement,ation issues including* the NRC* EAL Frequently Asked Questions WAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels Page 3 of 270
Grand GUif Nuciear Station EAL Basis Document Revision XXX
' ..*).*. . :<: }fi<}i;<1"t\iJ.;:i'/
for Non-Passive Reactors," November 2012 (ref. 4.1.1 ), GGNS conducted an EAL implementation upgrade project that produced the EALs discussed herein. 1
- 2.2 Fissio~. Prpd.LJct ~~JJ;iers Fission producf baftierth*~~~holds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that. define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Rotential Loss" signify the relative damage ~nd threa\ of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies a greater probability of barrier loss and reduced certainty *at maintaining the barrier. The primary fission product barriers are: A. Fuel Clad Barrier (FCB): *The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and _including the* isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria 1 The following criteria are the bases for event classification related to fission product barrier loss or potential loss:
- Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency:* Loss of any two barriers and loss or potential loss of the third barrier Page 4 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 2A . , EAL Organization The GqNs EAL scheme includes the following features:
- Division of the EAL set into three broad groups:
o EALs applicable under any plant operating modes - This group would be reviewed by th_e EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode. *
- o EALs applicable only-under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.
The purpose of the groups* is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-us.er for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
- Withinleach group, assignment of EALs to categor_ies and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The* GGNS EAL categories are aligned to and represent the NEI 99-01 "Recognition Categories." Subcategori~s are used in the GGNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The GGNS EAL categori~s and subcategories are listed below. The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information. Page 5 of 270.
Grand Gulf Nuclear Station EAL Basis Document Revision XXX EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I Any Operating Mode: A-Abnormal Rad Levels I Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1- Security Affecting Plant Safety 2- Seismic Event 3- Natural or Technological Hazard 4- Fire 5- Hazardous Gas 6- Control Room Evacuation 7- Emergency Director Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI) Hot Conditions: S - System Malfunction 1- Loss of ESF AC Power -- 2- Loss of Vital DC Power 3- Loss of Control Room Indications 4- RCS Activity 5- RCS Leakage 6- RPS Failure 7- Loss of Communications 8- Hazardous Event Affecting'Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown I Refueling System 1- RPV Level Malfunction 2- Loss of ESF AC Power 3- RCS Temperature 4- Loss of Vital DC Power 5- Loss of Communications 6- Hazardous Event Affecting Safety Systems Page 6 of 270
\ . ~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and -subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the
- emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
- 1. First character (letter): Corresponds to the EAL category as described above (A, C, E, F, Hor S)
- 2. Second character (letter): The emergency classification (G, S, A or U)
G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event
- 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1 ). If a
.category does not have a subcategory, this character is assigned the number one (1 ). .
- 4. Fourth character.(number): The numerical sequence of the EAL within the EAL subcategory. If the subcategor{has only one EAL, it is given the number one (1 ).
Classification (enclosed in rectangle):
. Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)
EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix. Page 7 of 270
)
- Grand Gulf Nuclear Stat.ion EAL Basis.Document Revision XXX Mode Applicability One or more of the.following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown,
- 5 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitiorils)
Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: An EAL basis section that provides GGNS-relevant inform_ation concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Reference(s}'. Source documentation from which the EAL is derived 2.6 Operating Mode Applicability 1 Power Operation Reactor is critical and the mode switch is in RUN 2 Startup lihe mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN and. average reactor coolant temp~rature is
>200°F .
4 Cold Shutdown The mode switch is in S~UTDOWN and average reactor coolant temperature is
\
s; 200°F 5 *Refueling The mode switch is in REFUEL or SHUTDOWN with one or more reactor vessel head closure bolts are less than fully tensioned DEF Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached'before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred. Page 8 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX ( 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition '(IC); This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the R~cognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds. EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top t9 bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices. 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes: after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG..:01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8). 3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, veriijcation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by 'plant personneL An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct
- observation by plant personnel, such that doubt ~elated to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing. radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. Page 9 of 270
~~Entergy. Grand Gulf Nuclear Station EAL Basis Document Revision XXX 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds .
an EAL does not warrant an emergency declaration provided that: 1) the activity pmceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or r conditions of this type may be subject to the reporting requirements of 10 § CFR 50. 72 (ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision) for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The E,mergency Director will need to determine if the effects or consequences of.the event orcohc:Htiori reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8). Page 10 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX . 3.2. 1 Classification of Multiple Events and Conditions
- When multiple emergency events or conditions are present, the user will identify* all met or exceeded EALs. The highest applicable EGL identified during this review is declared. For example: *
- If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency sh?uld be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
- If two Alert EALs are met, whether at one unit or at two units, an Alert shou.ld be declared.
3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and notwhen it was declared) .. Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated ~gainst the ICs and EALs applicable to the operating mode at the time of the new event or condition.
- For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.
3.2.3 Classific.ation of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short perio*d of time (Le., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to .all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
- Page 11 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 1* 3.2.4 Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition th.at meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met. 3:2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.
- 3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency dec,laration is warranted. In cases where no time-based criterion is specified, it is recognized thaf some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response~ In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures .. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An AJWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an.emergency declaration, then theclassifi~ation should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace peri.od" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision Page 12 of 270
\.
Grand Gulf Nuclear Station EAL Basis Document Revision XXX is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made \n the emergency classification process .. In these cases,* no emergency declaration is warranted; however, the guidance contained in
- NUREG-1022 (ref. 4.1.3) is applicable. Specifically,,,the event should be reported to the NRC in accordance with 10 CFR § 50. 72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in.
accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3). . Page 13 of 270
~Enter:gy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
4.0 REFERENCES
4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During
- Quickly Changing Events, Febru~ry 2, 2007.
4.1.3 NUREG-1922 Event Reporting Guidelines: 10CFR50.72 and 50;73 4.1.4 1O § CFR 50. 72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 1O § CFR 50. 73 License Event Report System 4.1.6 GGNS Technical Specifications Table 1.1-1, Modes 4.1. 7 GGNS Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 GGNS Emergency Plan 4.1.10 GGNS UFSAR 9.1.4.2.10.4 Storage of Fuel at the Independent Spent Fuel Storage Installation (ISFSI) 4.1.11 GGNS UFSAR 9.1.4.2.10 Qescription of Fuel Transfer 14.1.12 SOPP-018-1 Shutdown Operations Protection Plan 4.1.13 1O-S-01-12 Radiological Assessment and Protective Action Recommendations 4.2 Implementing 4.2.1
- 1O-S-01-1 Activation of the Emergency Plan 4.2.2 NEI 99-01 Rev. 6 to GGNS EAL Comparison Matrix 4.2.3 GGNS EAL Matrix J
Page 14 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 5.0 DEFINITIONS, ACRONYMS & ABBREVIATION1S 5.1 Definitions (ref. 4.1.1 except as noted) ......... ' Selected terms used in Initiating Condition, Emergency Action Lever-statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have .* specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress, or h*ave occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage t9 site equipme.nt because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the *Holtec System Multi-Purpose Canister (MPC) (ref. 4.1.10). Containment Closure The actions taken to secure containment (Primary or7Secondary) and their associated structures, systems, and components as a functionai barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established (ref. 4.1.12). Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level.
- Emergency Classification Level (ECL)
One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or
.consequences, and (2) resulting onsite and offsite response actions. The emergency*
classification levels, in ascending order of severity, are:
- Unusual Event (UE)
- Alert
- Site Area Emergency (SAE)
- General. Ernergency (~E)
Page 15 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Explosion I A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurizatio_n. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. Fire. Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within th~ room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integ*rity or HOSTILE ACTION that results in an actua.l loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes
- . attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-bas~d EALs should be used to address such activities (i.e., this may include violent acts between individuals in th~ SECURITY OWNER CONTROLLED AREA(SOCA)). Hostile Force Orie or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons. capable of killing, maiming, or causing destruction. Page 16 of 270
.. . . .. ,~ *),~:'.'** :~.,:;s~~0,: .
c:PEntergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. lmpede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry .of personnel into the affected room/area (e.g., requi_ring use of
- protective equipment, such as SCBAs, that is not routinely employed).,
Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage ofspent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating .Condition (IC) An event or.condition that aligns with the definition of one of the four emergency classification levels by virtue.of the potential or actual effects or consequences. Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. (ref. 4.1.9). RCS Intact The RCS should be *considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway .Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway (ref. 4.1.11). Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits.
- Page 17 of 270
---Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Safety System I
A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION. Security Owner Controlled Area (SOCA) The SOGA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOGA is the area between the SOGA Fence and the PROTECTED AREA Boundary. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure -of or; (2) that prevent effective access to equipment needed for the 'protection of the public. Any rel~ases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY. Site Boundary That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor (ref. 4.1.13) Unisolable An open or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event Page 18 of 270
-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requi'ring offsite response or monitoring are expected unless
- further degradation of SAFETY SYSTEMS occurs.
Page 19 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Page 20 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 5.2 Abbreviations/Acronyms ~F .......................... ,............................................................................ Degrees Fahrenheit. 0
*************************************************************************************************************************** Degrees AB ...........................................................................................................Auxiliary Building AC ........................................................................................................ Alternating Current AOP ...................... '. .......................................................... Abnormal Operating Procedure APRM .................................................................................. Average Power Range Meter ARI .................................................. :............................................. Alternate Rod Insertion A TWS ...................................................................... Anticipated Transient Without Scram BWR ............................................................................................... Boiling Water Reactor BWROG .................................................................. Boiling Water Reactor Owners Group CDE ... :***********************************************************************************Committed Dose Equivalent CFR ..................................................................................... Code .of Federal Regulations CNB ................................................................................................... Containment Barrier CS ............................................. *.* ..................................................................... Core Spray CTMT ...................................*.................................. *.............................. ,......... Containment DEF .............. ,~ ....... *................................................................................................ Defueled OBA ............................................................................................... Design Basis Accident DC ..... :................................... .' .................................*.................................... Direct Current DIG ......................................................................................................... Diesel Generator EAL .......................................................................................... '. .. Emergency Action Level ECCS ............................................................................. Emergency Core Cooling System ECL. ..,............................................................................... Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure.
EPA .............................................................................. Environmental Protection Agency EPG ............................................................................... Emergency Procedure Guideline EPP ..........................*............... ,.................................. :.......... Emergency Plan Procedure ERO ........................................................................... Emergency Response Organization ESF ......................................................................................... Engineered Safety Feature FAA .................................................................................. Federal Aviation Administration FBI ......................................................................*............. Federal Bureau of Investigation FCB .......................................................*................................................. Fuel Clad Barrier FEMA. ........................ :..................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ........................................... :......................................................... General Emergency Page 21 of 270
~::w Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
. HCTL ..................... :...................................................... Heat Capacity Temperature Limit HPCS ....................................................................................... High Pressure Core Spray IC .......................................................................................................... Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20) ISFSl. ........................................................ ,.. Independent Spent Fuel Storage Installation Ken ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER. ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPCS ........................................................................................ Low Pressure Core Spray LRW........................................................................................................ Liquid Radwaste LWR ................................................................................................... Light Water Reactor MPC ................................... Maximum Permissible Concentration/Multi-Purpose Canister MPH ........................................................................................................... Miles Per Hour mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MSCRWL. .................. ,.................................... Minimum Steam Cooling RPV Water Level MSIV ....................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line MW ......*.............................................................................................................. Megawatt NEI ............................................................................................... Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NESP ................................................................... National Environmental *studies Project NORAD ................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event NPP ................................................................................................... Nuclear Power Plant NRC .......... ~ ............................-......................................... Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System OBE ...................................................................................... Operating Basis Earthquake ODCM ............................................................................. Offsite Dose Calculation Manual ONEP ................................................................................... Off-Normal Event Procedure ORO ................................................................. :............... Offsite Response Organization
- PA ............................................................................................................... Protected Area PAG ........................................................................................ Protective Action Guideline PB ........................ ,................................................................................*........... Pushbutton PCIS ..................................................................... Primary Containment Isolation System PRA/PSA ..................... Probabilistic Risk Assessment I Probabilistic Safety Assessment Page 22 of 270
Grand Gulf Nuclear Station EAL Basis Document RevisiorJ XXX PSIG ................................................................................ Pounds per Square Inch Gauge R ................................................................................................................*......... Roentgen RCB .......................................................... .\...................................................... :RCS Barrier RCIC ................................................................................. Reacto~ Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man* RETS ....................................................... HRadiological Efflu~nt Technical Specifications RHR .......................................................................................... ;.. Residual Heat Removal RPS................................ ......... ;........................................... Reactor Protection System RPT .............. :...,::>:... ... . .. : ............................ ; ........................ Recirculation Pump Trip RPV ................................................. :......................................... Reactor Pressure Vessel RWCU ...... ; .... , ............... :............................................................ '..Reactor Water Cleanup SAP ...................._................................................................... Severe Accident Procedure ( I SAR .................................. '. ............................................................ Safety Analysis Report
.(
SBO .....................................................*................................ *... :................ Station Blackout SCBA.: ..................................................................... Self-Contained Breathing Apparatus SOCA .............................................................................. Security Owner Controlled Area SPDS ............................................ ~.......... '. .................... Safety Parameter Display System SRO ............................ \ .............................................................. Senior Reactor Operator SRV ...................... _. .............................................................................. Safety Relief Valve SSE ....................................................................................... Safe Shutdown Earthquake TEDE ............................................................................*... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TSC ......................... , ................................................................ Technical Support Center UFSAR ................................................................... Updated Final Safety Analysis. Report USGS ............................................................................ United States Geological Survey I . I Page 23 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 6.0 GGNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a GGNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the GGNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
GGNS NEI 99-01 Rev. 6 Example EAL IC EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 Page 24 of 270
( Grand Gulf Nuclear Station EAL Basis Document Revision XXX
~ ~'.C"l.,'~'";,'#i,'N , ,';
\ *:'.5'i"c{ffJ..?-.f,.. GGNS NEI 99.-01 Rev. 6
- Example EAL IC EAL AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 '2 CA2.1 CA2 1 CA3.1 CA3 1 2.
CA6.1 CA6 1 1CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 2 EU1.1 EU1 1 -I FA1.1 FA1 1 FS1 .1 FS1 1 FG1 . 1 FG1 1 HU1.1 HU1 1, 2, .3, HU2.1 HU2 1 I Page 25 of 270
-c~- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX GGNS NEI 99-01 Rev. 6 Example EAL IC EAL HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1 2 I HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 I HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1 2, 3 I SU6.1 SU5 1 Page 26 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX GGNS NEI 99-01 Rev. 6 Example EAL IC EAL , SU6.2 SUS 2 SU7.1 SU6 1, 2, 3
- N/A SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SAS 1 SAS.1 1 SA9 1 SS1.1 SS1 1 SS2.1 SSS 1 SS6.1 SSS 1 SG1.1 SG1 1 SG1.2 SGS 1 Page 27 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Page 28 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category A - Abnormal Rad Levels I Rad Effluent EAL Group: ANY (EALs in this category are applicable 'to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require _offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
- Events of this ?ategory pertain to the following subcategories:
- 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiab.le limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above Classifiable limits.
- 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification. *
- 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
Page 29 of 270
-::::~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
AU1.1 Unusual Event Reading on any Table A-1 effluent radiation monitor> column "UE" for ~ 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6. 7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec CTMT Vent II) 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6.7E-2 Ci/sec Q) II) Radwaste Building Vent cu (!) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6.?E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent "C
*5 . 7.33E+5 cpm C" Radwaste --- ---- ----
- J Page 30 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
All Definition(s): VALID -An indication, report, orcondition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the* condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a potential reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).' It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
- Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases .. The occurrence of an extended, uncontrolled radioactive release to the environment i.s indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and. conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based qn effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Releases \should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings. to exceed 2 times the limit established by a radioactivity discharge permit. Such releases are typically associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). Esca.lation of the emergency classification level would be via' IC AA 1. Page 31 of 270
-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. IAS-04-1-01-017-1 Process Radiation Monitoring
- 2. Offsite Dose Calculation Manual
- 3. XC-01017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values
- 4. NEI 99-01 AU1 Page 32 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. EAL: AU1.2 Unusual Event Sa.mple analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for~ 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an .additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): None Basis: This IC addresses a potential reduction in'.*the level of safety of the plant as indicated by a low:. level radiological release that exceeds regulatory commitments for an extended periog of time (e.g., an uncontrolled release): It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclea( power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further,*there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the,environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events arid' conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. Page 33 of 270
~Entergy Grand Gu.If Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm.drains, heat exchanger leakage in river water systems, etc.).
Escalation of the emergency classification level would be via IC AA 1. Reference(s):
- 1. Offsite Dose Calculation Manual
- 2. NEI 99-01 AU1 Page 34 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1*- Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.1 Alert Reading on any Table A-1 effluent radiation monitor> column "ALERT" for ~ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the*specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs AA 1.1, As1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classification Thresholds Release Point GE
- SAE Alert UE 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6. 7E-2 OVsec SBGT A/B\.
6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6.7E-2 Ci/sec CTMTVent II) 0 5.1 E+1 Ci/sec 5.1E+O Ci/sec 5.1 E-1 Ci/sec 6. 7E-2 Ci/sec Cl) II) Radwaste Building Vent n:J C) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec .6. 7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent "C
*:;; 7.33E+5 cpm C" Radwaste --- ---- ----
- i Page 35 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check; or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous ~r liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of bo,th plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AS1. Reference(s):
- 1. IAS.-04-1-01-017-1 Process Radiation Monitoring
- 2. XC-Q1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Value's *
- 3. NEI 99-01 AA1 Page 36 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiologic~I Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 1O mrem TEDE or 50 mrem thyroid COE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses> 10 mremTEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
\,,
Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude _represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily *or appropriately classified on the basis of plant conditions
- alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose ls set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AS 1.
-Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AA1 Page 37 of 270
-~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effl~ent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:
AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. , Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addres&es a release of gaseous or liquid radioactivity that results i'n projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. Tre inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. This EAL is assessed per the ODCM (ref. 2) Escalation of the emergency classification level would be via IC AS1. Page 38 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference( s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. Offsite Dose Calculation Manual
- 3. NEI 99-01 AA1 ,
Page 39 of 270
*-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological EfflueQt Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose
- greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:
AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
- Closed window dose rates> 10 mR/hr expected to continue for~ 60 min.
- Analyses of field survey samples indicate thyroid COE> 50 mrem for 60 min. of inhalation. I
( (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded . . Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis:* This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safe.ty of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
- The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE.
\
Page 40 of 270
~Entergy . Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
- Escalation of the emergency classification level would be via IC AS 1.
Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AA1 1.
Page 41 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document-Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:
AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor> column "SAE" for ~ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15_ minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-t* Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE I I I I I I 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6.7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec CTMT Vent II)
- l 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6. 7E-2 Ci/sec Cl)
II) Radwaste Building Vent
~
(!) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6. 7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6. 7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent
- E 7.33E+5 cpm
- l C" Radwaste --- ---- ----
- i Page 42 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: All Definition(s): I VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an . instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence,* or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.
- Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid- COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1. Reference(s):
- 1. IAS-04-1-01-017-1 ProcessRadiation Monitoring'
- 2. XC-Q1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values
- 3. NEI 99-01 AS1 Page 43 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses> 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorolo_gy are available. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AG1. Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AS1 Page 44 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
~
- Category: . A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:
AS1.3 Site Area Emergency Fi~ld survey results indicate EITHER of the following at or beyond the SITE BOU.NDARY:
- Closed window dose rates > 100 mR/hr expected to continue for~ 60 min.
- Analyses of field survey samples indicate thyroid COE> 500 mrem for 60 min. of inhalation.
(Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center
- of the reactor.
Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes path monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events ar:,d conditions. r The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AG1. Page 45 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Rec(?mmendations
- 2. NEI 99-01 AS1 Page 46 of 270
\Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:
AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor> column "GE" for ~ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected ar;,d the release start time is unknown, assume that the release . duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. f Table A-1 Effluent Monitor Clc1ssification Thresholds I Release Point I GE I SAE I Alert I UE I 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6. ?E-2 Ci/sec SBGT A/B
~ .
6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec CTMT Vent Cl) 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6. ?E-2 Ci/sec Cl) Cl) Radwaste Building Vent ca C) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6. 7E"'2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6. 7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent
":i 7.33E+5 cpm O" Radwaste
- J Page 4 7 of 270
-=::=w Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite* doses greater than or equal to the EPA Protec~ive Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.' The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thymid COE was ~stablished in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.
- Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.
Reference(s):
- 1. IAS-04-1-01-017-1 Process Radiation Monitoring
- 2. XC-Q1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values
- 3. NEI 99-01 AG1 Page 48 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A'- Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:
AG1.2 General Emergency . Dose assessment using *actual meteorology indicates doses> 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: . The pre-calculated effluent monitor values presented in EALs AA1 .1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. *
- Mode Applicability:
All Definition(s): . SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius fro.m the center of the rea9tor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PA<3s). It includes both m.onitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
- Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions .
alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses ,:.tt)~;::s,pectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the '1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Reference(s):
- 1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AG1 Page 49 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1 .3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
- Closed window dose rates> 1,000 mR/hr expected to continue for~ 60 min.
- Analyses of field survey samples indicate thyroid COE> 5,000 mrem for 60 min. of inhalation.
(Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): _ SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
- The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.
Page 50 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 1O~S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. t;JEI 99-01 AG1 Page 51 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by Fuel Pool Drain Tank low water level alarm, visual observation, water level drop in Upper Ctmt Pools, Aux Bldg Fuel Pools or the Fuel Transfer Canal AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
*
- Ctmt 209 Airlock (1021 K630)
- Ctmt Fuel Hdlg Area (1021 K626)
- Aux Bldg Fuel Hdlg Area(1021 K622)
Mode Applicability: All Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY- Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not includin~ the reactor vessel, comprise the refueling pathway. Basis:. This IC addresses a drop in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level drop will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations. Page 52 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX . Attachment 1 Emergency Action Level Technical Bases The effects of planned evolutions should be considered. For E?Xample, a refuelin-g bridge area . radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of Water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Categ~::>ry C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AA2. Reference(s):
- 1. 05-1-02-11-8 High Radiation During Fuel Handling
- 2. 04-1-01-021-1 Area Radiation Monitoring System
- 3. UFSAR 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation
- 4. NEI 99-01 AU2
/
\ Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
( Category: A - Abnormal Rad Levels I Rad Effluent* Subcategory: 2 - Irradiated Fuel Event Initiating* Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPG). IMMINENT - The trajectory of events or conditions is such" that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. '\) REFUELING PATHWAY- Reactor cavity (well), upper Gontainmentpool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the REFUELING PATHWAY. These events present radiological safety challenges to plant personnel and are precursors to a release of.radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask,causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1. This EAL escalates from AU2.1 in that the loss of level, in the affected portion of the
- REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include director indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based* on the totality of available indications, reports and observations.
While an area radiation monitor could detect a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings Page 54 of 270
--~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the. reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AS1. Reference(s): ~ 1. NEI 99-01 AA2 Page 55 of 270
---- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity* AND VALID alarm on any of the following radiation monitors:
- Ctmt Vent (P601-19A-G9)
- FH Area Vent (P601-19A-C11)
- Ctmt 209 Airlock (P844-1A-A 1)
- Ctmt Fuel Hdlg Area (P844-1A-A3)
- Aux Bldg Fuel Hdlg Area (P844-1A-A4)
Mode Applicability:
.I All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the
- condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
- Basis:
This EAL addresses events that have caused actual damage to an irradiated fuel assembly. These events* present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL *applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1. Page 56 of 270
~~~:Entergy -, Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 .Emergency Action Level Technical Bases This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
\
Escalation of the emergency classification level would be via IC AS1. Reference(s):
- 1. 05-1-02-11-8 High Radiation During Fuel Handling
- 2. 04-1-01-021-1 Area Radiation Monitoring System
- 3. UFSAR 12.3.4 Area Radiation and Airborne Radibactivity Monitoring Instrumentation
- 4. Offsite Dose Calculation Manual
- 5. NEI 99-01 AA2 Page 57 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 .;... Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 193 ft. (Level 2) on G41 R040A(B) Mode Applicability: All DefinitioQ(s): None Basis: This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel p~ol water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the eme'rgency classification level would be via IC AS 1 or AS2. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 193 ft. 2 1/a in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (LeveL3 - 183 ft. 2 1/ 8 in. rounded to 183 ft. for readability) (ref. 1). G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the-Control Room. The display cabinets are located in the 148' Control Building in the Lower Cab!e Spreading Room. Reference(s):
- 1. 05-S-01-FSG-011 Alternate SpentFuel Pool Makeup and Cooling
- 2. NEI 99-01 AA2 Page 58 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
\
Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 183 ft. (Level 3) on G41 R040A(B) Mode Applicability: All Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel pamage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized th~t this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC AG1 or AG2. Post-Fukushima order EA-12-051 required the installation of reliable SFP .level indication capable of identifying normal level (Level 1), SFP level 10 ft. above.the to.J:?;:PfJ,he fuel racks (Level 2 ~ 193 ft. 2 1/ 8 in. rounded to 193 ft. for readability) and SFP fevei alfffftop of the fuel racks (Level 3 - 183 ft. 2 1/ 8 in. rounded to 183 ft. for readability) (ref. 1). G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the148' Control Building in the Lower Cable Spreading Room. r Reference(s):
- 1.
- 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling
- 2. NEI 99-01 AS2 Page 59 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX ,
Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent
- Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannof be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 183 ft. (Level 3) on G41 R040A(B) for
~ 60 min. (Note 1)
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. ' Mode Applicability: All Definition(s): None
..~I Basis: I This EAL addresses a significant loss of spent fu~I pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well ~fter another General Emergency IC was met; however, it is included to provide classification diversity. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 1O ft. above the top of the fuel racks (Level 2 - 193 ft. 2 1/ 8 in. rounded to 193 ft. for readability) and SFP level .at the top of the fuel racks (Level 3 - 183 ft. 2 1/8 in. rounded to 183 ft. for readability) (ref. 1). . G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the148' Control Building in the Lower Cable Spreading Room. Reference(s):
- 1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling
- 2. NEI 99-01 AG2 Page 60 *of 270
~Entergy Grand 'Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical BasE:s Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
AA3.1 Alert Dose rate >. 15 mR/hr in EITHER of the following areas:
- Control Room (SD21-K600)
- Central Alarm Station (by survey)
Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: Areas that meet this threshold include the Control Room (CR) and the Central Ala-rm Station , (CAS). The Control Room is monitored for excessive radiation by one.detector, SD21-K600
- (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. While the CAS is in the Control Room Envelope, there are no permanently installed area radiation monitors 1in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area.
I This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary* to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level wo~ld be via Recognition Category A, C or F ICs. Page 61 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 06-IC-1021-R-1001 Area Radiation Monitoring Calibration
- 2. NEI 99-01 AA3 Page 62 of ,270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
AA3.2 Alert* An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table A-3 Safe Operation &( Shutdown Rooms/Areas Room/Area Mode Control Building 111' SWGR Rms (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A 103) 3 Auxiliary Building 93' RHR B Pump Room (1A 105) 3 Auxiliary Building 93' Corridor (1A 101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203) 3 Auxiliary Building 119' RHR B Pump Room (1A205) 3 Auxiliary Building 119' RCIC Room (1A204) 3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR B Room (1A306, 1A307) 3 Radwaste Building 118' Radwaste Control Room (OR241) 3 Mode Applicability: 3 - Hot Shutdown
')
Page 63 of 270
I
/
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilita.te entry of personnel into the* affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
- Ba~is:
This IC addresses elevated radiation levels fn certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performfng actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be
- _applicable.
For AA3.2, an Alert declaration isrwarranted if entry into the affected room/area is, or may be, proceduraily required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e:g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond' normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply:
- The plant is in an operating mode different than the mode specified for the affected room/area (Le., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not requJre entry into the affected room until Mode 3.
- The higher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g.,
radiography, spent filter or resin transfer, etc.).
- The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.
Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Page 64 of 270
~:=w Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown anti shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
EAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3 (Mode 3 only). Reference(s):
- 1. Attachment 2 Saf~ Operation & Shutdown Areas Tables A-3 & H-2 Bases
- 2. NEI 99-01 AA3 '
Page 65 of 270
-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
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Attachment 1 Emergency Action Level Technical Bases . Category C - Cold Shutdown I Refueling System Malfunction EAL Gro"up: Cold Conditions (RCS temperature s 200°F); EALs in this category are applicable only in one or more cold operating modes .. Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refueling, DEF - Defueled). The events of this category pertain to the following subcategories:
- 1. RPV Level RPV water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
- 2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV ESF buses.
- 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions.
- 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and e!llergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
- 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within.or external to the plant warrant emergency classification.
Page 66 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
- 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may'result in VISIBLE DAMAGE to or*
degraded performance of safety systems warranting classification. Page 67 of 270
~~*Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:
CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time li.mit is exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis:. Grand Gulf is equipped with multiple RPV water level instruments including: Wide Range, Fuel Zone, Shutdown Range, Upset Range, and Narrow Range (ref. 1). Multiple instruments on different reference and variable legs should be monitored. The Upset Range and Shutdown Range instruments share a common reference leg; therefore, Narrow Range instruments should be routinely monitored when relying on Shutdown or Upset Range instrument as the primary indication. With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 11.4 in. (ref. 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. With the plant in Refueling mode, RPV water level is normally maintained at or above the
- reactor vessel flange. Technical Specifications require at least 22 ft 8 in. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations (ref. 3). The RPV flange is at approximately 212 in. on the Shutdown Range. (ref. 4).
This EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent Page 68 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL recognizes that the minimum required RPV level can change several times during the\ course of a refueling outage as different plant configurations and system lineups are im'plemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be,'maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The '15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. Continued loss of RCS inventory may result in e~calation to the Alert emergency classification level via either IC CA1 or CA3. Reference{s):
- 1. 02-S-01-40 EP Technical Bases
- 2. 05-S-01 ~EP-2 RPV Control
- 3. Technical Specifications 3.9.6
- 4. 07-S-14-413 RPV Disassembly
- 5. NEI 99-01 CU1 Page 69 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:
CU1.2 Unusual Event RPV water level cannot be monitored AND EITHER
- UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory
- Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps/Pool
- Drywell equipment drain sump i
- Drywell floor drain sump
- CTMT equipment drain sump
- CTMT floor drain sump
- Suppression Pool
- RHR A, B, C, HPCS, LPCS, RCIC room sumps
- Auxiliary Building floor drain sump Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): UNf.SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. 'UNPLANNED-. A parameter changes or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode; the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate dip instrument which is re-spanned and re-scaled to indicate water level (ref. 1). Page 70 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be *evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and *calculating leakage from the RPV (ref. 2, 3). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses extemal to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory. that is available to keep the core covered. This EAL addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1 ). Sump and/or tank level changes ml:lst be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1H13-P680 Alarm Response Instruction Panel 1H13-P680
- 4. 05-S-01-EP-4 Auxiliary Building Control
- 5. NEI 99 CU1 Page 71 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
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Attachment 1 Emergency Action Level Technicai Bases
\
Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.1 Alert Loss of RPV inventory as *indicated by RPV water level < -42 in. (Level 2) Mode Applicability: 4 - Cold Shutdown, 5 "'.""" Refueling Definition(s): None Basis: The threshold RPV water level of -42 in. is the Level 2 actuation setpoint for HPCS and RCIC. I Although RCIC cannot restore RPV inventory in the cold condition, the L_evel 2 actuation setpoint is operationally significant and is indicative of a loss of RPV inventory significantly below the low RPV water level scram setpoint specified in CU1 .1 (ref. 1, 2). This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery. Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Retiloval suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):
- 1. Technical Specifications Table 3.3.5.1-1, Emergency Core Cooling. System Instrumentation
- 2. 04-1-02-1 H13-P601-16A-A4 Alarm Response Instruction Panel 1H13-P601 panel 16A-A4 RX LVL 2 (-42") LO
- 3. NEI 99-01 CA 1 Page 72 of 270
Grand Gulf Nuclear Station EALB:a'§jgt58@Gment Revision xxx Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level lniti~ting Condition: Significant Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for~ 15 min. (Note 1) AND EITHER
- UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory
* - Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Table C-1 Sumps/Pool
,
- Drywell equipment drain sump
- Drywell floor drain sump
- CTMT equipment drain sump
- CTMT floor drain sump
* . Suppression .Pool
- RHR A, B, C, HPCS, LPCS, RCIC room sumps
- Auxiliary Building floor drain sump Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. , UNPLANNED - A parameter change or an event that is not 1) the result of an *intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 73 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooli~g water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. I r 2, 3). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be irdicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could. be indicative of a loss of RPV inventory., This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial .reduction in the level of plant safety. ( For this EAL, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage . from the RPV.
- The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.
If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02,.1H13-P680 Alarm Response Instruction Panel 1H13-P680
- 4. 05-S-01-EP-4 Auxiliary Building Control
- 5. NEI 99-01 CA 1 Page 74 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction
\
Subcategory: 1*- RPV Level Initiating Condition_: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .1 Site Area Emergency CONTAINMENT CLOSURE not e$tablished AND RPV water level < -150 in. (Level 1) Mode- Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is- established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The threshold RPV water level of -150 in. is the low-low-low ECCS actuation setpoint (Level 1). The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further lowering of RPV water level and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier (ref. 1, 2). This IC addresses a significant and prolonged loss of RPV inventory- control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. - Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or ve"rifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory cont~bl
~age 75 of 270
Grand* Gulf Nuclear Station EAL Basis Docume_nt Revision XXX Attachment 1 Emergency Action Level Technical Bases functions. The difference in the specified RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s):
- 1. Technical Specifications Tabl~ 3.3.5.1-1, Emergency Core Cooling System Instrumentation
- 2. 04-1-02~1H13;;.P601-17~-E2 Alarm Response Instruction Panel 1H13-P601 panel 17A-E2 RX LVL 1 (-150'.') LO
- 3. NEI 99-01 CS1 Page 76 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
/
Attachm.ent 1 Emergency Action Level Technical Bases
- Category: C--: Cold Shutdown I Refu~ling System Malfunction
- Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:
CS1 .2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV water level< -167 in. (TAF) Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission . product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of-167 in.), core uncovery starts to occur (ref. *1 ). This IC addresses a significant and prolonged loss of RPV l~vel control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions*-needed for protection ofthe public and thus warrant a
. Site Area Emergency declaration."
Following an extended loss -of core decay heat removal and inventory mak~up, decay heat will cause reactor coolant boiling and a further reduction in reactor ve.ssel level. If RPV level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. Page 77 of 270
\
. Grand Gulf Nuclear Statio? EAL Basis Document Revision XXX Aftachment 1 Emergency Action Level Technical Bases This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues,~ NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in' the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the emergency classification level would be via IC CG1 or AG1. Reference( s):
- 1. 02-S-01--40 EP Technical Bases
- 2. NEI 99-01 CS1 Page 78 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: c*- Cold Shutdown I Refueling System ~alfunction Subcategory: 1 - RPV Level ( Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .3 Site Area Emergency RPV level cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any of the following:
- UNPLANNED rise in Suppression Pool level of sufficient magnitude to indicate core uncovery
- Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
- Containment/Drywell High Range Area Radiation Monitor (1 D21-K648B-C)
~ 100 R/hr Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. - UNISOLABLE -An open or breqched system line that cannot be isolated, remotely or locally. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown* mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which' is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must be evaluated against other potential Page 79 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1. Emergency Action Level Technical Bases sources of leakage such as cooling water sources inside the' drywell to ensure they are indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in Suppression Pool water level. could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS *that cannot be isolated could be.,indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core wiU rise, with corresponding indications on area radiation monitors. 100 R/hr is used for this indication on Containment High Range Radiation Monitors (1 D21-K648B and C). These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity elevation. This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of c9re decay heat removal and inventory makeup, decay heat will 1 cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. In this EAL, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other. potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Nlanagement.
- Escalation of the emergency classification level would be via IC CG1 or AG1 Page 80 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1 H 13-P680 Alarm Response Instruction Panel 1H13-P680
- 4. 05-S-01-EP-4 Auxiliary Building Control
- 5. 06-IC-1021-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration
- 6. NEI 99~01 CS1
- 7. Calculation J-021-1, Set Points Determination For High Range DW & Containment Radiation Monitors (021 System)
Page 81 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level lnitiaUng Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:
CG1 .1 General Emergency RPV level< -167 in. (TAF) for~ 30 min. (Note 1) AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-2 Containment Challenge Indications
- CONTAINMENT CLOSURE not established (Note 6)
- Drywell or containment hydrogen concentration> 4%
- UNPLANNED rise in containment pressure
- Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)
Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Page 82 of 270
-~~ Entergy Grand Gulf Nuclear Station *EAL Basis Document Revision XXX Attachment 1 Emergency Action*Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: When RPV level drops below -*167 in., core uncovery starts to occur (ref. 1). Four con.ditions are associated with a challenge to Containment integrity:
- CONTAINMENT CLOSURE is not established.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup* due to a core uncovery could result in an explosive. mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it" is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition.
Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 2).
- Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
- Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE.The MAX SAFE radiation levels are indicative of problems in the secondary containme-nt that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 3).
This IC addresses the inability to restore and maintain reactor vessel level above the top of *\ active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-Page 83 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases established prior to exceeding the 30-minute time limit, then .declaration of a General.
Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain. a containment I
hydrogeh':,gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s):
- 1. 02-S-01-40 EP Technical Bases
- 2. BWROG Emergency Procedure and Severe Accident Guidelines, Revision 3, p. B-16-64
- 3. 05-S-01-EP-4, Auxiliary Building Control
- 4.
- NEI 99-01 CG1 Page 84 of 270
~=--Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Teyhnical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment*
challen~ed EAL: CG1 .2 General Emergency RPV level cannot be monitored for~ 30 min. (Note 1) AND Core tincovery is indicated by any of the following:
- UNPLANNED rise in Suppression Pool level _of sufficient magnitude to indicate core uncovery
- Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery Containment/Drywell High Range Area Radiation Monitor (1 D21-K648B-C)
~ 100R/hr AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-2 Containment Challenge Indications*
- CONTAINMENT CLOSURE not established (Note 6)
- Drywell or containment hydrogen concentration > 4%
- UNPLANNED rise in containment pressure
- Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)
Page 85. of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. *
)
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be tntact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re~spanned and re-scaled to indicate water level (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must be evaluated against other potential sources of l.eakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in Suppression Pool water level could be indicative of RHR valve misalignment or" leakage. If the m9 ke-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise, with corresponding indications on area radiation monitors. 100 R/hr is used for this indication on Containment High Range Radiation Monitors (1 D21-K648B and C). These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity _elevation. Page 86 of 270
.~-~ .Entergy Grand Gulf Nuclear station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Four conditions are associated with a challenge to Containment integrity:
- CONTAINMENT CLOSURE is not established.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition.
Following ignition, hydrogen is consumed through formation pf diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 4).
- Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pr~ssure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
- Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of.a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 5).
This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual o~ IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CO~TAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment Page 87 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or . water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown . andLow-Power Operation at Commercial Nuclear Power Plants in the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1 H 13-P680 Alarm Response Instruction Panel 1 H 13-P680
- 4. BWROG Emergency Procedure and Severe Accident Guidelines, Revision 3, p. 8-16-64
- 5. 05-S-01-EP-4, Auxiliary Building Control
- 6. 06-IC-1021-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration
- 7. NEI 99-01 CG1
- 8. Calculation J-021-1, Set Points Determination For High Range OW & Containment Radiation Monitors (021 System)
Page 88 of 270
*~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 - Loss of ESF AC Power Initiating Condition: Loss of all but one AC power source.to ESF buses for 15 minutes or longer EAL:
CU2.1 Unusual Event AC power capability, Table C-3, to DIV I and DIV II ESF 4.16 KV buses reduced to a single power source for~ 15 min. (Note 1)
- AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time.limit is exceeded.
Table C-3 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
- DIV I DG (DG 11)
- DIV II DG (DG 12)
Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s): SAFETY SYSTEM.- A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor 6oolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a *safe shutdown condition; Page 89 of 270
(
~~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the greater time available to restore another power source to service . . Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognize_d in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.
- A loss of all offsite power with 1
a concurrent failure of all but one emergency ESF power source (e.g., an onsite diesel generator).
- A loss of all offsite power and loss of all emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being back-fed from the
,unit main generator.
- A loss of emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude. transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. This EAL is the cold condition equivalent of the hot condition EAL SA 1.1. Reference(s):
- 1. UFSAR Figure 8.1-001 Mairi One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section 8A Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 CU2 Page 90 of'270
~~~ Entergy Grand Gulf Nucl'ear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 ,.... Loss of ESF AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer EAL:
CA2.1 Alert Loss of all offsite and all onsite AC power to DIV I and DIV 11 ESF 4.16 KV buses for~ 15 min. (Note 1) Note 1: The Emergency Director shou.ld declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems
- classified as safety-related (as defined in 10CFR50.2):
Tho~e structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: ' This IC addresses a total *loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure cc;mtrol, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other powe'r sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may(be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of s'upplying power for long term decay heat removal systems. In particular, \suppression pool cooling systems would be essential subsequent to a station* blackout. Page 91 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore an ESF bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures .and pressures in v~_rious plant systems. Thus, when. in these r1199:!~.,J:his condition represents an actual Or"'lj'otential substantial degradation oftheo:levefcfj:t§:afety of the plant. . Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of the hot condition EAL SS1 .1. Reference{s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR s*ection 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section 8A Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 CU2 Page 92 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating C~ndition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Co,ntainment integrity is .established. UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant-response to a transient. The c~use of the parameter change
- or event may be known or unknown.
Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F) (ref. 1, 2). In the absence of reliable RCS temperature indication, classification is based on the concurrent loss of reactor vessel level indications per EAL CU3.2. This IC addresses an UNPLANNED. rise in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduc~d since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel Page 93 of 270
~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid rise in. reactor coolant temperature depending on the time after
- shutdown.
Escalation to Alert would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s):
- 1. Technical Specifications Table 1..1-1
- 2. 03-1-01-3 Plant Shutdown
- 3. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal
- 4. NEI 99-01 CU3 Page 94 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature
- EAL:
CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded,* or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. , Mode Applicability: 4 - Cold Shutdown, 5- Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA 1 based on an inventory 'loss or IC CA3 based on exceeding plant configuration-specific time criteria. Page 95 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Actio*n Level Technical Bases Reference{s):
- 1. 02-S-01-40 EP Technical Bases
- 2. Technical Specifications Table 1.1-1
- 3. 03-1-01-3 Plant Shutdown
- 4. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Rem9val
- 5. NEI 99-01 CU3
\
Page 96 of 270
~~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold ~hutdown I Refueling System Malfunc~ion Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED rise in RCS temperature to> 200°F for> Table C-4 duration (Note 1)
- OR UNPLANNED RPV pressure rise > 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Table C-4 RCS Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status Intact N/A 60 min.* established 20 min.* Not intact
) not established O min.
- If an RCS heat removal system is in operation within this time frame and *RCS temperature is being reduced, the EAL is not applicable. '
Mode Applicability: 4- Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, syst~ms, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 97 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
( Basis: In the absence of reliable RCS temperature indication, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4. This EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the 1.evel of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.* The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantial degradation in plant safety. Finally, in the case where there is a rise in RCS temperature, the RCS is not intact and CONTAINMENT CLOSURE is not established, no heat-up durationis'allowed (i.e., O minutes). This is because 1) the evaporated .reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The RCS pressure rise threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or AS1 . . Reference(s): .. 1 .* Technfd~fl~fJecifications Table 1.1-1 . 2. 03-1-01-3 Plant Shutdown
- 3. 04-1-01-E12-2 Shutdown Cooling and Alternat~ Decay Heat Removal
- 4. NEI 99-01 CA3 Page 98 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes*or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 voe on required vital 125 voe buses 11 DA and 11 DB for~ 15 min. {Note 1) Note, 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during ,and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis Vital DC buses 11 DA and 11 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1. 72 volts/cell. These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC) (ref. 1, 2). This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions raise the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if I Page 99 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event.. A loss of Vital DC power to Train A would not warrant an* emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA 1 or CA3, *or an IC in Recognition Category A. This EAL is the cold condition equivalent of the . hot condition EAL 882.1. Reference(s):
- 1. Calculation No: EC-01111-14001 Station Divi~ion I Battery 1A3 and Division II Battery 183 Discharge Capacity during Extended Loss of AC Power
- 2. UFSAR 8.3.2.1.1 Station DC Power
- 3. NEI 99-01 CU4
~age 100 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Cgndition: . Loss of all onsite or offsite communications capabilities. EAL: CUS.1 Unusual Event Loss of all Table C-5 onsite communication methods. OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods
*State/
System Onsite NRC Local Station Radio System x GGNS ,Pla.nt Phone System x Public Address System x Emergency Notification-System (ENS) x Commercial Telephone System x x Satellite Phones x x Operational Hotline I'* x Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s):
- None Page 101 of 270
~*Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of o,n-site or offsifecommunications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the N RC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriff's Department, Louisiana Department of Environmental Quality, Tensas Parish Sheriff's Office, and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness.
- The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
This EAL is the cold. condition equivalent of the hot condition EAL SU7.1. Reference(s):
- 1. GGNS Emergency Plan Section 7.5; Communications Systems
- 2. 04-S-01-R61-1 Plant Communications
- 3. NEI 99-01 CU5 Page 102 of 270
~-~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
- Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
- Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table C-6 Hazardous Events
- Seismic event (earthquake)
- Internal or external FLOODING event
- High winds or tornado strike
- FIRE
- EXPLOSION
- Other events with similar hazard characteristics as determined by the Shift Manager Page 103 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required .if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it irJ a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or ~eliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second , SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; Page 104 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or
- reliability of the SAFETY SYSTEM train.
VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE shouid be significant enough to cause concern regarding the operability or reliability of the $AFETY SYSTEM train. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equJvalent of the hot condition EAL SA8.1. Reference(s):
- 1. EP FAQ 2016-002
- 2. NEI 99-01 CA6 Page 105 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials . associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. The GGNS ISFSI is located wholly within the plant PROTECTED AREA Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs. Page 106 of 270
~~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1 .1 Unusual Event
\
Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack) 1
> EITHER of the following:
_)
- 60 mrem/hr (gamma + neutron) on the top of the overpac.k
- 600 mrem/hr (gamma+ neutron) on the side of the overpack (excluding inlet and outlet ducts)
Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUND~RY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is. sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values. The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Cate.gory A IC AU1, is used here to distinguish between non-emergency and emergency conditions (ref. 2). The emphasis for this classification is the, _degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose Page 107 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency AcUon Level Technical Bases rate*. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. ( Security-related events for ISFSls are covered under ICs HU1 and HA1. Reference(s):
- 1. UFSAR 9.1.4.2.10.4 Storage of Fuel at the lndeper:,dent Spent Fuel Storage Installation
- 2. GGNS HI-STORM 100 10 CFR 72.212 Evaluation Report Licensing Basis Document, Revision 10, Section 4.2.4 (Section 5. 7) Radiation Protection program
- 3. NEI 99-01 E-HU1 Page 108 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachmen~ 1 Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained _intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: * \ A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCS}: The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves.
- C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the
. containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an a
Alert to Site Area Emergency or a General Emergency. The EALs in this categoryt require evaluation of the loss and potential loss thresholds listed in the fission product barrier rnatrix of Table F-1. "Loss" and "Potential Loss" signify the relative . damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of an) two barriers 0 General Emergency: Loss of any two barriers and loss or potential loss of third barrier The logic u~ed for emergency classification based on fission product barrier monitoring srould reflect the following considerations:
- The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
Page 109 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
- Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
- For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded.
I
- The fission product barrier thresholds specified within a scheme. r~f!,ect plant-specific GGNS design and operating characteristics. ** * **_M~ ,
- As used in this category, the term RCS leakage encompasses not just those types qefined 1n Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.
The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.
- At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more ._assurance that there was no immediate need to escalate to a General Emergency.
Page 110 of 270
~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
.)
Attachment 1 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1) (I Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown. Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss 'of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1 Reference(s):
- 1. NEI 99-01 FA 1 Page 111 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases I Category: F - Fission Product Barrier Degradation Subcategory: NIA Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency
~oss or potential loss of any two barriers (Table F-1)
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 ~ Hot Shutdown Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is Weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
- One barrier loss and a second barrier loss (i.e., loss~ loss)
- One barrier .loss and a second barrier potenti~l loss (i.e., loss - potential loss)
- One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier lbss thresholds in addition to offsite dose assessments would require cc;mtinual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have - greater assurance that escalation to a General Emergency is less IMMINENT. Reference(s):
- 1. NEI 99-01 FS1 Page 112 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1) ( Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown
/
Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
- Loss of Fuel Clad, RCS arid Containment Barriers
- Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
- Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
- Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference( s):
- 1. NEI 99-01 FG1 Page 113 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Techn_ical Bases 1
Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the t.hreshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are: A. RPV Water Level B. RCS Leak Rate C. Containment Conditions D. Containment Radiation I RCS Activity E. Containment Integrity or Bypass F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.*
- Thresholds are assigned sequential numbers within each barrier column beginning with 1
number one (ex., FCB1, FCB2 ... FCB6). If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promote? a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user. determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Page 114 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FGt 1, FS1 .1, and FA 1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then B, ... , F. Page 115 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB)
Category Loss Potential Loss Loss Potential Loss Loss Potential Loss FCB2 RPVwater level cannot be RCB1 RPV water level cannot be A FCB1 SAP entry is required restored and maintained restored and maintained None None CNB1 SAP entry is required RPVWater > -167 in. (TAF) > -167 in. (TAF) Level or cannot be determined or cannot be determined RCB2 UNISOLABLE break in any of the following:
.. Main steam line RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:
CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER: B .. RCIC steam Line One or more EP-4 radiation One or more EP-4 MAX None RCS Leak Rate None None RWCU Feedwater HPCS
. Operating Limits One or more EP-4 area temperature Operating
. SAFE area radiation levels One or more EP-4 MAX SAFE area temperature RCB3 Emergency Depressurization Limits levels is required CNB5 Containment pressure > 15 psig CNB3 UNPLANNED rapid drop in containment pressure following CNB6 Drywell or containment c None None RCB5 Drywell pressure > 1.23 psig None containment pressure rise hydrogen concentration > 4%
CTMT due to RCS leakage CNB4 Containment pressure CNB7 Parameters cannot be restored Conditions response not consistent with and maintained within the safe LOCA conditions zone of the HCTL curve (EP Figure 1) FCB3 Containment radiation (RITS-D K648B or C) > 400 R/hr RCB6 Drywell radiation (RITS-K648A CNB8 Containment radiation (RITS-CTMT Rad/ None None None FCB4 Primary coolant activity or D) > 100 R/hr K648B or C) > 7000 R/hr RCS > 300 µCi/gm dose Activity equivalent 1-131 CNB9 UNISOLABLE direct downstream pathway to the E environment exists after CTMT None None None None Containment isolation signal None Integrity or CNB101ntentional Containment Bypass venting per EPs F FCB5 Any condition in the opinion of the Emergency Director FCB6 Any condition in the opinion of the Emergency Director that RCB7 Any condition in the opinion of the Emergency Director that RCB8 Any condition in the opinion of the Emergency Director that CNB11 Any condition in the opinion of the Emergency Director that CNB12Any condition in the opinion of the Emergency Director that Emergency that indicates loss of the fuel indicates potential loss of the indicates loss of the RCS indicates potential loss of the indicates loss of the indicates potential loss of the Director clad barrier fuel clad barrier barrier RCS barrier Containment barrier Containment barrier Judgment Page 116 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold: I FCB1 SAP entry is required Definition(s): None Basis: Emergency Procedures (EPs) specify entry to the Severe Accident Procedures (SAPs) when core cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2). The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad. This threshold is also a Potential Loss of the Containment barrier (CNB1 ). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCB1 ). SAP entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. The Loss threshold represents the EOP requirement for entry into the SAPs. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. EP FAQ 2015-004
- 4. NEI 99-01, RPV Water Level Fuel Clad Loss 2.A Page 117 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold: FCB2 RPV water level cannot be restored and maintained> -167 in. (TAF) or cannot be determined . Definition(s): None Basis: An RPV water level instrument reading of-167 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier. When RPV water level cannot be determined, EPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary m~ans of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EP-5 specify these means, which include emergency depressurization of the. RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. In high-power A TWS/failure to scram events, EOPs may direcUhe operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification.
, Page 118 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachme'nt 1 - Emergency Action Level Technical Bases This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold RCB1. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressuri.zation ofthe RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also: specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In . some events, elevated. RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to a.ssess the capability of low-pressure injection sources to restore RPV water level' or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minirtnize loss of RPV inventory. The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an ev;aluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation 'below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad b~rrier is specified. Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. 05-S-01-EP-2A A TWS RPV Control 4 NEI 99-01 RPV Water Level Potential Loss 2.A Page 119 of 270
(
Grand Gulf Nucl~ar Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: I None Page 120 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: I None Page 121 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Loss Threshold: I None _Page 122 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: I None Page 1.23 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Loss Threshold: FCB3 Containment radiation (RITS-K648B or C) > 400 R/hr Definition(s): None Basis: The containment radiation monitor reading (425 R/hr rounded to 400 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131.
- Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 1.6% fuel clad damage (ref. 1). Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold RCB6 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency. There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation I RCS 1 Activity. Reference(s):
- 1. XC~Q1021 .. 17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values
- 2. 04-1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A Page 124 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX ,-
Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Loss Threshold: FCB4 Coolant activity> 300 µCi/gm dose equivalent 1-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for. iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation I RCS Activity. Reference(s): .
- 1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A Page 125 of 270
I Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Potential Loss Threshold: None Page 126 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: I None Page 127 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier:* Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
I None Page 128 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1*- Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: FCB5 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad Barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 129 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment
. Degradation Threat: Potential Loss Threshold: FCB6 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad Barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 130 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases. Barrier: Reactor Coolant System Category: A. RPV Water Level Degradation Threat: Loss Threshold: RCB1 RPV water level cannot be restored and maintained> -167 in. (TAF) or cannot be determined Definition(s): None. Basis: An RPV water level instrument reading of-167 in. indicates level is at the top of active fuel (TAF) (ref. 1). TAF is sJgnificantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previo'usly required isolation of the RCS and Containment barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of
- reversing the lowering level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA.. By definition, a LOCA event is a Loss of the R(?S barrier.
When RPV water level cannot be determined, EOPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). The instructions in EP-5 specify emergency depressurization of the RPV, which is defined to be a Loss of th~ RCS barrier (RCS Loss RCB3).
- Note that EP-2A, A TWS RPV Control, may require intentionally lowering RPV water level to
-167 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - RPS Failure EALs.
- In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or 886.1 will dictate the need for emergency classification.
- This water level corresponds to the top of active fuel and is used in the EOPs to indicate a*
challenge to core cooling. Page 131 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases The RPVwater level threshold is the same as Fuel Clad barrier Potential Loss threshold FCB2. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restori,ng RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, ,this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term, "cannot be restored and maintained above," means the value of RPV water .level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the, RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. There is no RCS barrier Potential Loss threshold associated with RPV Water Level. Reference( s): , 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. 05-S-01-EP-2A A TWS RPV Control
- 4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 132 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RPV Water Level D.egradation Threat: Potential Loss Threshold:
I None Page 133 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
- Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss 1 Threshold:
RCB2 UNISOLABLE bre'ak in any of.trn~hfollowing:
- Main steam line
- RCIC steam line
- RWCU
- Feedwater
- HPCS Definition(s):
UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but .in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires-the normal process flow of a system outside containment (e.g., EP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see Loss CNB9) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected tc;> the RCS.
- Even though the High Pressure Core Spray (HPCS) injects into the RCS, it is included in this EAL due to the potential for an inter-system loss of coolant back flowing from the discharge Page 134 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases lines (via failed isolation valves and check valves) and out through a break in the piping. A HPCS failure that does not result in back flow of RCS and out through a break should not be considered as meeting the EAL threshold. Large high-energy lines that rupture outside containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, remotely or locally,, the RCS barrier Loss threshold is met. Reference(s):
- 1. NEI 99-01 RCS Leak Rate RCS Loss 3.A Page 135 of 270
. -=::.::w Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 ..:.. Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:
RCB3 .Emergency Depressurization is required Definition(s): None Basis: Emergency Depressurization in accordance wit~ the EOPs (ref. 1, 2) is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. EP-2 RPV Control - Emergency Depressurization allows terminating the depressurization if necessary to maintain RCIC as an injection source. This would require closing the SRVs. Even though the SRVs may be reclosed, this threshold is still met due to the requirement for an Emergency Depressurization having been met (ref. 2).
- Reference(s):
- 1. 05-S-01-EP-2 RPV Control - Emergency Depressurization
- 2. EP FAQ 2015-003
- 3. NEI 99-01 RCS Leak Rate RCS Loss 3.B Page 136 of 270
\\ ~*.Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX l Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:
- One or more EP-4 area radiation Operating Limits
- One or more EP-4 area temperature Operating Limits Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of UNISOLABLE primary system leakage outside the containment. The EP-4 entry condition values define this RCS threshold because they are the Operating Limits (maximum normal operating values) and signify the o,nset of abnormal system operation. When parameters reach this level, equipment failure or mis~operation may be
- occurring. Elevated parameters may also adversely affect the ability to gain. access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1).
In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area radiation condition doe.s not necessarily indicate that a primary system is discharging into the Auxiliary .Building since this may be
- caused by radiation shine from nearby steam lines or the movement of radioactive materials.
Conversely, *a high area radiation condition in conjunction with other indications (e.g. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discha.rging into the Secondary Containment. Potential loss of RCS based on primary system leakage outside the containment is determined from EOP temperature or radiation EP-4 Operating Limits (Max Normal Operating values) in Page 137 of 270
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside containment.
An EP-4 Operating Limit value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control 1 systems functioning properly. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a reduction in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by EP-4 Operating Limit values escalates to a Site Area Emergency when combined with Containment Barrier Loss thresholds CNB 2 or CNB9 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. Reference(s):
- 1. 05-S-01-EP-4 Auxiliary Building Control
- 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 138 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C. CTMT Conditions Degradation Threat: Loss Threshold: RCB5 Drywell pressure > 1.23 psig due to RCS leakage Definition(s): None Basis: The drywell high pressure scram setpoint is an entry condition to EP-2, RPV Control, and EP-3, Containment Control (ref. 1, 2). Normal containment pressure control functions (e.g., operation of drywell and containment cooling, vent using containment vessel purge, etc.) are specified in EP-3 in advance of less desirable but more effective functions (e.g., operation of containment sprays, etc.). In the design basis, containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a hi.gh-energy release into the containment for which normal pressure control systems are inadequate or incapable ofreversing the rising pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control containment venUpurge (ref. 3). The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect containment pressure. Drywell pressure greater than 1.23 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.23 psig should not be considered an RCS barrier Loss.
- The 1.23 psig value is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.
There is no RCS barrier Potential Loss threshold associated with CTMT Conditions.* ( Page 139 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX r , Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-3 Containment Control
- 3. UFSAR Section 6.2.1, Containrnent Functional Design
- 4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A Page 140 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: I None Page 141 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:
RCB6 Drywell radiation (RITS-K648A or D) >. 100 R/hr
- Definition(s):
None Basis: The drywell radiation monitor reading (150 R/hr rounded to 100 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits (ref. 1). This value is lower than that specified for Fuel Clad Barrier Loss threshold FCB3 since it indicates a loss of the RCS Barrier only. There is no RCS barrier Potential Loss threshold associated with CTMT Radiation/ RCS Activity. Reference(s):
- 1. XC-Q1021-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values *
- 2. 04-1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 142 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 143 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
I None Page 144 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: I None Page 145 of 270
~-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1- Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:
RCB7 Any condition in the opinion of the Emergency Director that indicates loss of the RCS Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A
- Page 146 of 270
~Entergy. Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:
RCBS Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in , determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference{s):
- 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 147 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Loss Threshold: I None Page 148 of 270
Grand Gulf Nuclear Station EAL Basis Document Rev_ision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold: I CNB1 SAP entry.is required Definition(s): None ) Basis: EPs specify entry to the SAPs when core cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPVinjection sources when level cannot be determined (ref. 1, 2). The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad. This threshold is also a Loss of the Fuel Clad barrier (Loss FCB1 ). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (Loss RCB1 ). SAP entry, therefore, represents a Loss of two barrlers and a Potential Loss of a third, which requires a General Emergency classification. The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold FCB1. The Potential Loss requirement for entry into the SAGs indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs specify the conditions when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure ade,guate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and greater potential for containment failure. In conjunction with; the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold *results in the declaration of a General Emergency. There is no Containment barrier Loss threshold associated with RPV Water Level. Page 149 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 1ttachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. EP FAQ 2015-004
- 4. 'NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 150 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:
- One or more EP-4 MAX SAFE area radiation levels
- One or more EP-4 MAX SAFE area temperature levels Definition(s):
FLOODING - A condition where water is entering a room or area fasJerthqJ\Jnstalled equipment is capable of removal, resulting in a rise of water level within the room or area. UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the containment. The MAX SAFE values define this Containment barrier threshold because they are indicative of problems in the Secondary Containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1). In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area temperature condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by a fire or loss of area cooling. Conversely, a high area temperature condition in conjunction with other 'indications (e.g. room FLOODING, high area radiation levels, reports of steam in the secondary containmenh:,an unexpected rise in feedwater flowrate, or unexpected , main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. The Max Safe area temperature values and the Max Safe area radiation values are each the highest value of these parameters at. which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the Page 151 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions undeJwhich RPV depressurization is required. There is no Containment barrier Potential Loss threshold associated with RCS Leak Rate. Reference(s):
- 1. 05-S-01-EP-4 Auxiliary Building Control
- 1. NEI 99-01 RCS Leak Rate PC Loss 3.C J
Page 152 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: I None Page 153 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: 'C. CTMT Conditions Degradation Threat: . Loss Threshold:
CNB3 UNPLANNED rapid drop in containment pressure following containment pressure rise Definition{s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Rapid UNPLANNED loss of containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity. This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Reference{s):
- 1. NEI 99-01 Primary Containment Conditions PC Loss 1.A Page 154 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Loss Threshold: CNB4 Co'ntainment pressure response not consistent with LOCA conditions Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter' change or event may be known or unknown .. Basis: Containment pressure should rise as a result of mass and energy release into the containment from a LOCA. Thus, containment pressure not rising under these conditions indicates a loss of containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indjcator for a containment bypass condition. Reference(s):
- 1. USAR Table 6.2-5, Summary of Short Term Containment Responses to Recirculation Une and Main Steam Line Breaks
- 2. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design
- 3. NEI 99-01 Primary Containment Conditions PC Loss 1.B L
Page 155 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment _Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNBS Containment pressure > 15 psig Definition(s): None Basis: When the containment pressure exceeds the maximum allowable value (15 psig) (ref. 1), containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). This pressure is based on the containment design pressure as identified in the accident analysis. If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an
- unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.
The threshold pressure is the containment internal design pressure. Structural acceptance testing demonstrates the capability of the containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.
- Reference(s):
- 1. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design
- 2. 05-S-01-EP-3 Containment Control
- 3. NEI 99-01, Primary Containment Conditions PC Potential Loss 1.A Page 156 of 270
.Grand Gulf Nuclear"Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold:
CNBS Drywell or containment hydrogen concentration> 4% Definition(s): 'None Basis: In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest conc~ntration that can ~upport ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 1). If hydrogen concentration reaches or exceeds the lower flammability limit 1 as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the containment, loss of the Containment barrier could occur. Reference(s):
- 1. 02-S-01-40 EP Technical Bases (EP-3 step H-3)
- 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.8 Page 157 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment* Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNB7 Parameters cannot be restored and maintained within the safe zone of the HCTL curve (EP Figure 1) Definition(s): None Basis: The Heat Capacity _Temperature Liniit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
- Suppression bhamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
- Suppression chamber pressure above Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. The term "cannot be restored and maintained within" means the parameter value(s) is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to the parameter value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained within a specified limit does not require immediate action simply because ther current value is outside the limit, but does not permit extended operation outside the limit; the threshold must be considered reached as soon as it is apparent that operation within the limit cannot be attained. Page 158 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C Page 159 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold: I None Page 160 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: CNB8 Containment radiation (RITS-K648B or C) > 7,000 R/hr Definition(s): None Basis: In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (Loss RCB6) and a loss of the Fuel Clad barrier (Loss FCB3) have already occurred. This threshold, therefore, represents a General Emergency classification. The containment radiation monitor reading (7,350 R/hr rounded to 7,000 R/hr for readability) corresponds to an instantar;ieous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed (ref. 1). This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. There is no Containment barrier Loss threshold associated with CTMT Radiation/RCS Activity. Reference(s):
- 1. XC-Q1021-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values
- 2. 04--1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 NEI 99-01 Primary Containment Radiation Potential Loss 4.A Page 161 of 270 (
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment
\
Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB9 .) UNISOLABLE direct downstream pathway to the environment exists after Containment isolation signal Definition(s): UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release* pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity. This threshold also applies to a containment bypass due to a HPCS or LPCS line break outside containment with injection check valve failure allowing an UNISOLABLE direct \pathway for RCS release to the environment. The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not. protected by the PrimarY Containment Isolation System (PCIS). Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE main steam line or RCIC steam line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the' main condenser is available with an UNISOLABLE main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the Page 162 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action* Level Technical Bases environment. These minor releases are assessed using the Category A, Abnormal Rad Rel.ease I Rad Effluent, EALs.
- The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
EP-3 Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a VALID containment isolation signal, the Containment barrier should be considered lost. Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Reference(s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 163 of 270
~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment **
Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB10 Intentional Containment venting per EPs Definition(s): None Basis: EP-3, Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded. The threshold is met when the operator begins venting the containment in accordance with 3, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 1). Intentional venting of containment for containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Reference(s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 CTMT Integrity or Bypass Containment Loss 3.8 Page 164 of 270
Grand Gulf Nuclear-Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMTfr1tegrity or Bypass Degradation Threat: Potential Loss Threshold: I None Page 165 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:
CNB11 Any condition in the opinion of the Emergency Director that indicates loss of the Containment Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Reference( s):
- 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A.
Page 166 of 270
-c;:=--Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:
CNB12
- Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment Barrier Definition(s):
None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whetheUhe Containment Barrier is p*otentially lost. The Emergency Director should also consider whether or not to declare the barrier.potentially lost in the event that barrier status cannot be monitored.
- Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 167 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
- 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physica! control of the plant.
- 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
- 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
- 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown
- 5. Hazardous Gas Toxic, .corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. *
- 6. Control Room Evacuation Events that are indicative- of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
- 7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms o*r events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.
Page 168 of 270 .
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat
. EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by GGNS Security Shift Supervision I OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands wiU be met by the station.
, I HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve a~ end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).
OWNER CONTROLLED AREA (OCA) - For the purposes of classification, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOCA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern .for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which- access is controlled. Page 169 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) T~e capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result _ in potential offsite exposures. SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA 1 and HS1. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards ContingencyPlan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold addresses the receipt of a credible securi_ty threat. The credibility of the threat is assessed in accordance with the Security Plan for GGNS. The third threshold addresses the threat from the ~pact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with 11-S-82-1 Security Contingency Events (ref. 2). Page 170 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Emergency plans and* implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for- GGNS (ref. 1). Escalation of the emergency classification level would be via IC HA 1. Reference{s):
- 1. GGNS Security Plan
- 2. 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HU1 Page 171 of 270
~Entergy Grand Gul,f Nuclear Station EAL Basis Document Revision XXX-Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: . HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by GGNS Security Shift Supervision
-OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:
All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station, HO~TILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, orwater using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or
- felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).
HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA - For the purposes of classffication, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOCA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Page 172 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:. This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the , need to prepare the plant and staff for a potential aircraft impact. . Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. ) Security plans and terminology are based bn the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
- As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response ,
Organizations (OROs), a*llowing them to be better prepared should it be necessary to consider further actions. This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. The second threshold addresses the threat from the impact of an aircraft on the plant, and the . anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with 11-S-82-1 Security Contingency Events (ref. 2). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, shoulq not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or I Page 173 of 270
-===~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 . ;. . Emergency Action Level Technical Bases threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1).
Escalation of the emergency classification level would be via IC HS1. Reference(s):
- 1. GGNS Security Plan
- 2. 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HA1 Page 174 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:
HS1 .1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by GGNS. Security Shift Supervision Mode Applicability:
- All Definition(s):
HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA - For the purposes of classification, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOCA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controll~d. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Page 175 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security~related event (ref. 1., 2). Security plans and terminology.are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This EAL does not apply to incidents that are accidental everJts, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security.:.sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1). Reference(s):
- 1. GGNS Security Plan
- 2. 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HS1 Page 176 of 270
~~Entergy Grand Gulf. Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety*
Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: Unusual Event Seismic event> OBE as indicated by annunciation of EITHER of the following on SH13P856:
- Containment Operating Basis Earthquake (P856-1A-A3)
- Drywell Operating Basis Earthquake (P856-1 A-AS)
Mode Applicability: All Definition(s): None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthqua~e greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., perform walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of
- the level of safety of the plant.
Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the U.S: Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, pryclude a timely emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst Page 177 of 270
. . . . . .. .:. *.. */ ,.\jf/~¥~ti,,
Grand Gulf Nuclear Station EAL Basis Document Revisi'6rY'XXX Attachment 1 - Emergency Action Level Technical Bases with the following GGNS coordinates: 32° O' 27" north latitude, 91° 2' 53" west longitude. (ref. 2). Alternatively, near real-time seismic activity can be accessed via the NEIC website. Reference(s):
- 1. 05-S-02-Vl-3 Earthquake
- 2. UFSAR 2.1.1 Site Location and Description
- 3. NEI 99-01 HU2 Page 178 of 270
I
-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
- Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTE;D AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation
- of the level of safety of the plant.
This EAL addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. / If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1. A tornado *striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending -from the base of a thunderstorm.
- Reference(s):
- 1. 05-1-02-Vl-2 Hurricanes, Tornados and Severe Weather
- 2. NEI 99-01 HU3 Page 179 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level .Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability: All . Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability-to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, Sor C. Page 180 of 270
~.Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Refer to EAL CA6.1 or SA8.1 for internal FLOODING affecting more than one SAFETY SYSTEM train.
Reference(s):
- 1. 05-1-02-Vl-1 Flooding
- 2. NEI 99-01 HU3 Page 181 of 270
Grand Gulf Nuclear Station EAL Basis Docurnent Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: ,* H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3~3 Unusual Event Movement of personnel .within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) i Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s):
- 1. NEI 99-01 HU3 Pag~ 182 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
- t*tl.Ji4.:~--, :,t~~runusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff fro\m accessing the site via personal vehicles (Note 7)
Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s):
- 1. NEI 99-01 HU3 Page 183 of 270
Grand Gulf Nuclear Station EAL Bas'is Document Revision XXX Attachment 1 - Emergency Action Level Technical BJses Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety pf the plant I EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):
- Report from the field (i.e., visual observation)
- Receipt of multiple (more than 1) fire alarms or indications
- Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 115 minutes to declare after the time limit is exceeded.
Table H-1 Fire Areas
- Unit 1 Containment
- Unit 1 Auxiliary Building
- Unit 1 Turbine Building
- Control Building
- Diesel Generator Rooms
- SSW Pump & Valve Rooms Mode Applicability:
All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 184 of 270
""~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. ' Upon receipt, operators will take *prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purpose~, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. Depending upon the plant mode at the time of the event, escalation of the emergency classification levelJ would be via EAL CA6.1 or SAS.1. The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or fiel~ validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Table H-1 Fire Areas a're those areas,that contain equipment necessary for safe operation .and shutdown of the plant (ref. 1, 2). * ** Reference(s):
- 1. 05-S-02-V-1 Response to Fires
- 2. 1O-S-03-2 Response to Fires
- 3. NEI 99-01 HU4 Page 185 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) (Note 11) AND 1:"he fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 11: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the containment or drywell. Table H-1 Fire Areas
- Unit 1 Containment
- Unit 1 Auxiliary Building
- Unit 1 Turbine Building
- Control Building
- Diesel Generator Rooms
- SSW Pump & Valve Rooms Mode Applicability:
All - This MUST be all modes because only the containment and drywell are excluded in modes 1 and 2, but other areas are not Definition(s): FIRE- Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID -Ah indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condi.tion's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 186 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment Emergency Action Level Technical Bases Basis:
This IC_ addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
- This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be ,indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time rs allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual Fl RE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. This EAL is not applicable for the containment or drywell in Modes 1 and 2. The air flow design and TS requirements for operation of Containment Fan Coolers and the drywell cooling system are such that multiple detectors would be expected to alarm for a fire in the containment or drywell. A fire in the containment or drywell in these modes would therefore be classified under EAL HU4.1. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the 1 report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix Ato this part specifies that "Structures, syste_ms, and r components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
- Page 187 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases I In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. The 30 minute requirement begins upon receipt of a single VALID fire detection system alarm: The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15 minute requirement beginning with the verification of the fire by field report. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2). Reference(s):
- 1. 05-S-02-V-1 Response to Fires
- 2. 1O-S-03-2 Response to Fires
- 3. UFSAR Appendix 9A Fire Hazard Analysis Report
- 4. NEI 99-01 HU4 Page 188 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX v Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level df safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has
- been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Mode Applicability: All Definition(s): FIRE - Combustion characterized qy heat and light. Sources of smoke such as slipping drive 1 belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is .controlled. ' Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially d~grade the level of plant safety. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. Reference(s):
- 1. NEI 99-01 HU4 Page 189 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentialJjt:rdegradin*g . the level of. s~f~t}l::6{:,ttt~.:~Plant
. . .. : .. *~
.: .. ,,'; ' ....\.\- ; **.. ~
EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. Reference(s):
- 1. NEI 99-01 HU4 Page 190 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX I Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for* normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table H-2 Safe Operation & Shutdown Rooms/Areas
*Room/Area Mode Control Building 111' SWGR Rms (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A 103) 3 Auxiliary Building 93~ RHR B Pump Room (1A 105) 3 Auxiliary BuHding 93' Corridor (1A 101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203). 3 Auxiliary Building 119' RHR B Pump Room (1A205) 3 Auxiliary Building 119' RCIC Room (1A204) \.
3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR B Room (1A306, 1A307) 3 Radwaste Building 118' Radwaste Control Room (OR241) 3 Mode Applicability: 3 - Hot Shutdown Page 191 of 270
~~Entergy* Grand Gulf Nuclear Station EAL Basis Document Revision XXX .
Attachment 1 - Emergency Action Level Technical Bases Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition 'represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating. experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply:
- The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.
- The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system'testing).
- The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.
- If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. Page 192 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an I action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). EAL HA5.1 mode applicability has been limited to the mode limitations of Table H-2 (Mode 3 only). Reference(s):
- 1. Attachment 2 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
- 2. NEI 99-01 HA5 Page 193 of 270
-=::.=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H -, Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel Mode Applicability: All Definition(s): None Basis: This IC addresse.s an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on ..shift personnel., Activation of the ERO and emergency response facilities will assist in responding to these challenges. Transfer of plant control begins when the last licensed operator leaves the Control Room. Escal_ation of the emergency classification level would be via IC HS6 .
../
Reference(s):
- 1. 05-1-02-11-1 Shutdown from the Remote-Shutdown Panel
- 2. NEI 99-01 HA6 Page 194 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document RevisionXXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:
HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel ' AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1):
- Reactivity (Modes 1 and 2 only)
- RPV water level
- RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded: or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Mode Applicability: 1 -Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown 1 locati6 n(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC FG1 or CG1 Page 195 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-1-02-11-1 Shutdown from the Remote Shutdown Panel
- 2. EP FAQ 2015-014
- 3. NEI 99-01 HS6 Page 196 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
. Attachment 1 - Emergency Action Level Technical Bases Category: H- Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL:
HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring ,offsite response or monitoring are expected unless* further degradation of SAFETY SYSTEMS occurs. Mode Applicability: All Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following ~esign basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis:
)
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT. Reference(s):
- 1. NEI 99-01 HU?
Page 197 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL:
HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent _may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-b 9 sed EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER.CONTROLLED AREA). OWNER CONTROLLED AREA - For the purposes of classification, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOCA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Page 198 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX _ Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant .* declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT. Reference(s):
- 1. NEI 99-01 HA?
Page 199 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL:
HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indjcate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead . to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, I.and, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA - For the purposes of classification, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOGA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
\
PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Page 200 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY. Reference(s):
- 1. NEI 99-01 HS7 Page 201 of 270
~Entergy. Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category:. H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL:
HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of plilysical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or. felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. OWNER CONTROLLED AREA - For the purposes of classification, the Security area between the OCA detection fence and the PROTECTED AREA boundary known as the Security Owner Controlled Area (SOCA) in the GGNS Emergency Plan.
- PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel s~fety.
( Page 202 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY. Reference(s):
- 1. NEI 99-01 HG?
_) Page 203 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-*related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories:
- 1. Loss of ESF AC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4. 16 KV .ESF buses.
(
- 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity: This category includes loss of power to or degraded voltage on the 125V DC vital buses.
- 3. Loss of Control Room Indications Gerta.in events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory. ,
- 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp I.
uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
- 5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core.
The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel
. clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
Page 204 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases
- 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, A TWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
- 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.
Page 205 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite AC power capability to ESF buses for 15 minutes or longer EAL:
SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 KV buses for~ 15 min. (Note 1) Note 1: , The Emergency Director should declare the event promptly upon determ,ining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
- DIV I DG (DG 11) .
- DIV II DG (DG 12)
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC ESF buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the ESF buses, whether or not the buses are powered from it. Page 206 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
. Attachment t - Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. '
Escalation of the emergency classification level would be via IC SA 1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electr,ic Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 SU1 Page 207 of 270
~Entergy Orand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction 1.. 1 Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all but one AC power source to ESF buses for 15 minutes or longer EAL:
SA1.1 Alert AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 KV buses reduced to a single power source .for~ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
- DIV I DG (DG 11 )
- DIV II DG (DG 12)
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Tho.se structures, systems and components that are relied upon to remain functional during* and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; Page 208 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases (2) The capability to shut down the reactor and maint:ain *it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.
- A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
- A loss of all offsite power and loss of all ESF emergency power sources (e.g., onsite
- diesel generators) with a single train of ESF buses being back-fed from the unit main generator.
- A loss of ESF emergency power sources (e.g., onsite diesel generators) with a single train of ESF emergency buses being fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of . power. Escalation of the emergency classification level would be via IC SS1. This EAL is the hot condition equivalent of the cold condition EAL CU2.1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 SA1 Page 209 of 270
\
/
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction SubGategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to ESF buses for 15 minutes or longer EAL: SS1 .1 Site Area Emergency Loss of all offsite and all onsite AC power to DIV I and DIV 11 ESF 4.16 KV buses for~ 15 min. (Note 1) ( Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition{s): SAFETY SYSTEM - A system required for safe. plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during . and following design basis events ta* assure: (1) The integrity of the reactor cool.ant pressure boundary; (2) The capability to shut down the reactor and maint,ain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink Mitigative strategies using other power sources (HPCS DIV 111 diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sour9es must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout.. In addition, fission product barrier monitoring capabilities may be degraded Page 210 of 270
. .~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases under these conditions. This IC represents a condition that involves actual or likely major failures of plantJunctions needed for the protection of the public.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent bf the cold condition EAL CA2.1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02~1-4 Loss of AC Power
- 6. NEI 99-01 SS1 Page 211 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to ESF buses EAL:
SG1 .1 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV 11 ESF 4.16 KV buses AND EITHER:
- Restoration of at least one ESF 4.16 KV bus in < 4 hours is not likely (Note 1)
- RPV water level cannot be restored and maintained> -191_ in.
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems ' classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite e~posures. Basis: Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-191 in.) (ref. 6). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling). This IC addresses a prolonged loss of all power sources to AC ESF emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat Page 212 of270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these * ! conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification fror:n Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC ESF emergency bus by the end of the analyzed station bl~ckout coping period. Beyond this time, plant responses1 and event trajectory are subject to greater uncertainty, and there is a greater likelihood of challenges to . multiple fission product barriers, 4 hours is the site-specific SBO coping analysis time (ref. 4). The estimate for restoring at least one ESF emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time
- available to prepare for, and implement, protective actions for the public.
The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section 8A Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. 02-S-01-40 EP Technical Bases
- 7. NEI 99-01 SGt Page 213 of 270
~Entergy Grand Gulf Nuclear-Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction I
' Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all ESF AC and vital DC power sources for 15 minutes or longer EAL: SG1 .2 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV 11 ESF 4.16 KV buses for ~ 15 min. (Note 1) AND Indicated voltage is < 105 voe on vital 125 voe buses 11 DA and 11 DB for
~ 15 min. (Note 1)
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown. Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result
- in potential offsite exposures.
Basis: Vital DC buses 11 DA and 111 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61. cells with a design minimum of 1. 72 volts/cell. These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC) (ref. 6, 7).
- This IC addresses a concurrent and prolonged loss of both emergency ESF AC and Vital DC power. A loss of all emergency ESF AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, Page 214 of 270
~~-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
Mitigative strategies using other pow~r sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these bus.es. These power sources must be controlled in* accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular,.suppression pool cooling systems would be essential subsequent to a station b,lackout. Aloss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency ESF AC and Vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
- Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1:-4 Loss of AC Power
- 6. Calculation No: EC-01111-14001 Station Division I Battery *1A3 and Division II Battery 183 Discharge Capacity during Extended Loss of AC Power
- 7. UFSAR 8.3.2.1.1 Station DC Power
- 8. NEI 99~01 SG8 Page 215 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:
SS2.1 Site Area Emergency Indicated voltage is < 105 voe on vital 125 voe buses 11 DA and 11 DB for .
- ?! 15 min. (Note 1)
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15
- minutes to declare after the time limit is exceeded.
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut downJhe reactor and main.tain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of acpidents which could result in potential offsite exposures .. Basis: Vital DC buses 11 DA and 11 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1. 72 volts/cell. These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to, 105 VDC) (ref. 1, 2). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In *modes above Cold Shutdown, thi~ condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition .EAL CU4.1. ,
/
Page 216 of 270
~ffl... Entergy!
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):I
- 1. Calculation No: EC-01111-14001 Station Division I Battery 1A3 and Division II Battery 183 Discharge Capacity during* Extended Loss of AC Power
- 2. UFSAR 8.3.2.1.1 Station DC Power
- 3. NEI 99-01 888
/
Page 217 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer
- EAL:
SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-2 Safety System Parameters
- Reactor power
- RPV water level
- RPV pressure
- Containment pressure
- Suppression Pool water level
- Suppression Pool temperature Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing ifin the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in. a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. UNPLANNED - A parameter change or ~n event that is not 1) the result'of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 218 of 270
-~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Controi Room. This condition is a precursor to a*more significant event and represents a potential degradation iri the level of safety of the plant.
- As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Cdntrol Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room.
I An event involving a loss of plant indications, ,annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significar;itly impaired the capability to perform emergency assessments. In particular, emergency assessments , necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. ' This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters _are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via EAL SA3.1. Reference(s):
- 1. UFSAR 7.5 Safety-Related Display Instrumentation 2 .. NEI 99-01 SU2 Page 219 of 270
~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S ~ System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: . UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1 Alert Ari UNPLANNED event re.suits in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) AND 1 Any significant transient is in progress, Table S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 I minutes to declare after the time limit is exceeded. Table S-2 Safety System Parameters 1*
- Reactor power
- RPV water level
** RPV pressure
- Containment pressure
,
- Suppression Pool water level
- Suppression Pool temperature
- Table S-3. Significant Transients
- Reactor scram
- UNPLANNED drop in reactor thermal power > 25%
- Electrical load rejection > 25%
electrical load
- ECCS injection
(
- Thermal power' oscillations > 10%
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Page 220 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those $tructures, systems and components that are relied upon to remain functional during and following design basis events to assure: * (1) The integrity of the reactor coolant pressure boundary; (2) The capabilitr to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown .. Basis: This IC addresses the difficulty *associated with monitoring rapidly changing plant conditions during a transientwithout the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the 1 plant.
- I, As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Roo1m sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot'be determined from the indications and recorders on a main control board, Page 221 of270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases
_the SPDS or the plant computer, the availability of other parameter values may be compromised as well.
- Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC FS1 or AS1. Reference(s):
- 1. UFSAR 7.3 Engineered Safety Features Systems
- 2. NEI 99-01 SA2 Page 222 of 270
;'~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL:
SU4.1 Unusual Event Offgas Pretreatment radiation monitor high-high alarm Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: The Offgas Pretreatment monitors radioactivity in the Offgas system downstream of the Offgas condenser. The monitor detects the radiation level that is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser. The Hi-Hi. al.arm, if alarming, indicates that the radioactivity present at the recombiner effluent discharge is at or above the Technical Specification 3. 7.5 limit of 380 millicuries per second of Noble Gases. (ref. 1) This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs. ( In the event that the Offgas Pretreatment Radiation Monitor High-High Alarm is out of service, the use of offgas flowrates and Offgas Pretreatment Radiation monitor readings is a viable contingency action to classify the EAL. See chart in 04-1-02-1 H13-P601-19A-D7, Alarm Response Instruction for OG PRE-TREAT RAD HI-HI alarm. Reference(s):
- 1. Alarm Response Instruction 04-1-02-1H13-P601-19A-D7
- 2. UFSAR 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
- 3. Technical Specification 3. 7.5 Main Condenser Offgas
- 4. 05-1-02-11-2 Offgas Activity High
- 5. NEI 99-01 SU3 Page 223 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category:
- S - System Malfunction Subcategory: 4- RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable
- limits (
- EAL:
SU4.2 Unusual Event Coolant activity> 0.2 µCi/gm dose equivalent 1-131 for> 48 hours OR Coolant activity> 4.0 µCi/gm dose equivalent 1-131 instantaneous Mode Applicability: 1 - Power Operation, 2 - Startup, 3 -. Hot Shutdown Definition(s): None Basi~: . This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs. Reference(s):
- 1. Technical Specification 83.4.8, RCS Specific Activity bases
- 2. UFSAR Section 15.6.4 Steam System Piping Break Outside Containment
- 3. NEI 99-01 SU3 Page 224 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual. Event RCS unidentified or pressure boundary leakage > 10 gpm for~ 15 min. (Note 1) OR RCS identified leakage > 25 gpm for~ 15 min. (Note 1) OR Leakage from the RCS to a location outside containment> 25 gpm for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. Identified leakage is leakage into the drywell, such as that from pump seals or valve pa'cking, that is captured and conducted to a collecting sump; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. Unidentified leakage is all leakage into t,he drywell that is not identified leakage (ref. 2, 3). Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2, 3). This IC addresses RCS leakage which may be a precursor to a more significant event. hi this case, RCS leakage has been detected ~nd operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS Page 225 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases
. )
mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions ~thus apply to leakage into the containment, or a location outside of containment. . The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, I therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of Recognition Category A or F. Reference(s):
- 1. UFSAR Section 5.2.5, Detection of Leakage Through Reactor Coolant Pressure Boundary
- 2. Technical Specification Definitions Section 1.1
- 3. Technical Specification 3.4.5
- 2. NEI 99-01 SU4 Page 226 of 270
~~-- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: , Automatic or manual scram fails to shut down the reactor EAL:
SU6.1 Unusual Event An automatic scram did not shut down the reactor as indicated by reactor power> 5% after any RPS setpoint is exceeded
. AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in shutting down the reactor as indicated by reactor power~ 5% (APRM downscale) (Note 8)
Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that res~lts in a reactor shutdown, and either a subsequent oper~tor manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
- The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).
A successful scram has occurred when there is sufficient rod in-sertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%. For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref.- 2). Page 227 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Following any automatic RPS scram signal, operating procedures (e.g., EP-2) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial-'failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event (ref. 3). Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to ~ 5% is not considered a successful automatic scram. If automatic initiation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. In the event that the operator identifies a reactor scram is IMMINENT and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following, indications that a scram setpoint was exceede~), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50. 72 should be considered for the transient event. Following the failure of an automatic reactor scram, operators will promptly initiate ma.nual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems .. Page 228 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX: Attachment 1 -: Emergency Action Level Technical Bases A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the , condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via EAL SA6.1. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
- If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
- If the signal generated"'as a result of plant work does not cause a plant transient ~nd the scram failure is deter:rnined through other means (e.g., assessment of test results), then
' this IC and associated EALs are not applicable and no classification is warranted.
Reference(s):
- 1. Technical Specification Table 3.3.1.1_-1 Reactor Protection System Instrumentation
- 2. 05-S-01-EP-2A ATWS RPV Control
- 3. 05-S-01-EP-2 RPV Control
- 4. NEI 99-01 SUS Page 229 of 270
~Entergy. Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:
SUS.2 Unusual Event 1 A manual scram did not shut down the reactor as indicated by reactor power> 5% after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in .shutting down the reactor . as indicated by reactor power::; 5% (APRM downscale) (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): , Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or COQlplete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shuttihg down the
- reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. I This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subseqLJent automatic or manual scram is successful in shutting down the reactor (reactor powers 5%) (ref. 1).
- Page 230 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM *downscale setpoint of 5%.
For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch-, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2). Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. Successful automatic or manual initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to fhe EAL threshold in the absence of any required subsequent manual scram actions. If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor p*ower below the power associated with the SAFETY SYSTEM design (~ 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1. Following the failure of an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions _are successful in shutting down the reactor, core heat gen'eration will quickly fall to a le~el within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch. Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is succe~sful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action 'does not include manually driving in control rods or implementation of boron injectioh strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the c.ontrol Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power' level prior to the event, availability of the conde,nser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator man~al actions taken at the reactor control consoles Page .231 of 270 iI
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the, plant conditions needed to meet either IC SA6 or FA 1, an Unusual Event declaration is appropriate for this event.
Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
- If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to
- automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
- If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.
Reference(s):
- 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation
- 2. 05-S-01-EP-SA A TWS RPV Control
- 3. 05-S-01-EP-2* RPV Control
- 4. NEI 99-01 SU5
( Page 232 of 270
~-Entergy Grand Gulf Nuclear Station EAL Basis, Document Revision XXX Attachment 1 . .;. Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic'or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1 Alert An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% AND Manual scram actions taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) are not successful in shutting down the reactor as indicated by reactor power> 5% (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the contr0I rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The irtegrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;r (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic;; or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an :actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action Page 233 of 270
Gra.nd Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases taken away from the reactor control consoles since this event entails. a significant failure of the RPS. This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by subsequent manual scram actions that fail* to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMSwere designed(> 5%). For the purposes of emergency classification, successful m.anual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2). For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to or below 5% is not considered a successful automatic scram. If automatic actuation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. rHowever, a successful automatic initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the, reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS6 or FS 1, an Alert declaration is appropriate for this event. _It is recognizeqJb.*~t,:PJ~.nt responses or symptoms may also require an Alert declaration in accordance with)tJ\~ii~~eognition Category F ICs; however, this IC and EAL are included to ensure a timelyemergency declaration. Page 234 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. ) Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation
- 2. 05-S-01-EP-2A A TWS RPV Control
- 3. 05-S-01-EP-2 RPV Control
- 4. NEI 99-01 SAS Page 235 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: i SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor as indicated by reactor power
>5%
AND All actions to shut down the reactor are not successful as indicated by reactor power> 5% AND EITHER: RPV water level cannot be restored and maintained > -191 in. OR Heat Capacity Temperature Limit (HCTL) exceeded (EP Figure 1) Mode Applicability: 1* - Power Operation, 2 - Startup Definition(s): / . \ SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to,prevent or mitigate the consequence$ of accidents which could resul~ Jf.l potential offsite exposures. *
.Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown; all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. Page 236 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases This EAL addresses the following:
- Any automatic reactor scram signal followed by subsequent manual scram actions that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1 ), and
- Indications that either core cooling is extremely challenged or heat removal is extremely challenged.
Reactor shutdown achieved by use of control rod insertion methods in EP-2A step Q-1 are also credited as a successful shutdown provided reactor power can be reduced to or below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1) The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers. Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 1). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence (ref 2). The Heat Capacity Temperature Limit (HCTL, EP Figure 1) is the highest suppression pool water temperatwe from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppression pool temperature. The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step of section SPT in EP-3, Containment Control, is reached (ref. 3). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. In some instances, the emergency classification resulting from this EAL may be hig,her than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in th~t th~ Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. Escalat;on of the emergency classification level would be via IC AG1 or FG1. Page 237 of 270
~==-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05~S-01-EP-2A, A TWS RPV Control
- 2. 05-S-01-EP-5, RPV Flooding
- 3. 05-S-01-EP-3, Containment Control
- 4. NEI 99-01 SSS Page 238 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
SU7.1 '
- Unusual Event Loss of all Table S-4 onsite communication methods 1
OR Loss of all Table S-4 State arid local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods
""' State/
System Onsite NRC Local Station Radio System x GGNS.Plant Phone System x Public Addmss System x Emergency Notification System (ENS) x Commercial Telephone System x x Satellite Phones x x Operational Hotline x Mode Applicability: 1 - Power Ope~ation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Page 239 of 270
Grand Gulf ~uclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriff's Department, Louisiana Department of Environmental Quality, Tensas parish Sheriff's Office, and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness. The third EAL condition ad9.-resses a tqtal loss of the communications methods used to notify the NRC of an emergency declaration.
- This EAL is the hot condition equivalent of the cold condition .EAL CU5.1.
Reference(s):
- 1. GGNS Emergency Plan Section 7.5, Communications Systems
- 2. 04-S-01-R61-1 Plant Communications
- 3. NEI 99-01 SU6
\
Page 240 of 270
Grand Gulf Nuclear Station EAL Basi~ Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction I Subcategory: _ 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:. SAS.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
- Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
- Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table S-5 Hazardous Events
- Seismic event (earthquake)
- Internal or external FLOODING event
- High winds or tornado strike
- FIRE
- EXPLOSION
.* Other even,ts with similar hazard _characteristics as determined by the Shift Manager Mode Applicability:
1 - *Power Operation, 2 - Startup, 3 - Hot Shutdown Page 241 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Definition(s):
EXPLOSION~ A rapid, violent and catastrophic failure of a piece of equipment que to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE- Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: I (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;
* (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readfly observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. I Basis:* This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has .indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Page 242 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation oJ th'e emergency classification level would be via IC ,FS1 or AS1. This EAL is the hot condition equivalent of the cold condition- EAL CA6.1. Reference(s):
- 1. EP FAQ 2016-002
- 2. NEI 99-01 SA9 Page 243 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
/
Background
NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to petform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air
. intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.
Page 244 of 270
/
--~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GGNS Table A-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:
IOI / 501 ACTIONS IOI 03-1-01-2 Power Operations LOWER Power by reducing Recirculation flow until 62.2% core flow (70 MCR 1 mlbm/hr) is reached.
- INSERT Control Rods per Control Rod Movement Sequence. MCR TECH SPEC TRIGGER (SR 3.3.2.1.2, SR 3.3.2.1.4) MCR 1 IF Reactor power has been reduced below the 1HPSP OR the LPSP, THEN PERFORM one of the following: Required Surveillanc.es or enter LCO for TS 3.3.2.1 CHECK OPEN the following valves on 1H13-P870-6C: - MCR 1
- a. N11-F029A, HP TURB EXTR To MSRA 1ST STG RHT
- b. N11-F029B, HP TURB EXTR To MSR B 1ST STG RHTIF N11-F029A OR N11-F029B are NOT open, THEN RETURN MSR 1ST Stage Reheaters to service per SOI 04-1-01-N11-1.
CHECK OPEN the following valves on panel 1H13-P870-6C: MCR 1
- a. N36-F01 OA, EXTR STM SPLY TO FW HTR SA
- b. N36-F01 OB, EXTR STM SPLY TO FW HTR 58
- c. N36-F011A, EXTR STM SPLY TO FW HTR 6A
- d. N36-F011 B, "EXTR STM SP4Y TO FW HTR 68 TAKE handswitches for the following valves to OPEN position on panel 1H13-P870-6C:
- a. N36-F013A, FW HTR SA EXTR STM BTV
- b. N36-F013B, FW HTR 58 EXTR STM BTV
- c. N36-:F012A, FW HTR 6A EXTR STM BTV
- d. N36-F012B; FW HTR 68 EXTR STM BTV NOTIFY the following of the power reduction: MCR
- Load Dispatcher (Woodla,nds)
* *Duty Manager (IF unexpected power reduction) * (SMEPA)(1-601-261-2318 OR 1-601-261-2313) * * (SMEPA) Site Representative
- Radwaste
- Radiation Protection
- Chemist-ry
* *NRC Resident Inspector Page 245 of 270
~~~* Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES
- These notifications Must be made by Shift Manager IF LP Turbine inlet temperature is >491°F, and N11-F028A and N11- TURB BLDG 1 Not F028B are open, THEN SIMULTANEOUSLY THROTTLE the following ELEV 133 Required valves on 1H22-P177 to CONTROL LP Inlet Temperatures within a AREA4 band of 470° F to 490° F while monitoring LP Turbine Inlet differential ROOM temperatures within 30° F ~(comparing A side to B side). 1T325
. .N11-F028A . N11-F028B *I IF LP Turbine inlet temperature is >491 °F, and N11-F028A and N11-F028B are closed, THEN SLOWLY, SIMULTANEOUSLY LOWER MSR-A/B HTG STM FEED CONT manual setpoint to CONTROL LP Inlet Temperatures within a band of 470° F to 490° F while monitoring LP Turbine Inlet differential temperatures within 30° F (comparing A side to B side).
LOWER Reactor power by INSERTING control rods to specified MCR 1 Control Rod in-sequence position per 17-S-02-400. At approximately 48% Reac;;tor power, PERFORM the following on MCR 1 panel 1H13-P601. VERIFY the following valves Open:
- B21-F033 INBD MSL DR SOL TO MN CNDSR
- B21-F069 OTBD MSL DR SOL TO MN CNDSR
- OPEN B21-F016 At approximately 50% Reactor Power, PERFORM the following:
SHUTDOWN 1 Reactor Feed Pump per SOI. 04-1-01-N21-1. VERIFY RFPT Bis operating normally on master controller. MCR 1 RAISE FW MASTER LVL CONT setpoint to approximately 39" MCR 1 TRANSFER the RFPT A SP CONT to MAN. MCR 1 -- SLOWLY LOWER speed of RR'PT A USING RFPT A SP CONT by MCR 1 DEPRESSING the OUT o pushbutton. OBSERVE speed of RFPT B raises to maintain RPV water level, OR control it manually FURTHER REDUCE speed of RFPT A using RFPT A SP CONT MCR 1 in MAN until it reaches low speed stop. TRANSFER speed control of RFPT A to SPEED AUTO by MCR 1 DEPRESSING the OBSERVE the FW AUTO pushbutton extinguishes AND the SPEED AUTO, RAISE, AND LOWER pushbuttons backlight. FURTHER REDUCE RFPT A speed using RFPT A LOWER MCR 1 pushbutton. WHEN RFPT A speed reaches 1100 rpm, THEN TRIP RFPT A by MCR 1 DEPRESSING the RFPT A MAN TRIP pushbutton CHECK F014A, RFP A DISCH VLV starts to close. MCR 1 Page 246 of 270
)
~Entergy_ Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES REOPEN F014A, RFP A DISCH VLV WHEN RFPT A coasts down to zero speed, THEN RESET turning gear MCR by pressing A TURN GEAR OPER RESET pushbutton.
OBSERVE turning gear engages automatically, unless RFPT A is rolling on min flow. IF turning gear fails to engage, THEN MANUALLY ENGAGE the TURB BLDG 1 Not turning gear locally by PRESSING DOWN the manual engaging lever. ELEV 133 Required AREA3 ROOM 1T307, 1T309 CHECK OPEN/OPEN the following Drain valves on 1H22-P175: N/A N/A These 1N11-F019A, RFPT A HP IN DRVLV steps are not 1N11-F023A, RFPT A HP IN DR VLV required 1N11-F018A, RFPT A IP IN DR VLV to be 1N11-F021A, RFPT A IP IN DR VLV performed 1N11-F042A, RFPT A IP IN DRVLV to Shut down and 1N33-F021A, RFPT A ABOVE SEAT DR Cool down
*1 N33-F022A, RFPT A ABOVE SEAT DR the plant.
1N33-F023A, RFPT A BELOW SEAT DR 1N33-F024A, RFPT A BELOW SEAT DR RETURN FW MASTER LVL CONT setpoint to approximately 36" MCR IF desired, RESET RFPT A trip using the RFPT A TRIP RESET MCR pushbutton SHUTDOWN 1 Circulating Wtr Pump per SOI 04-1-01-N71-1 CHECK that CTCS balls are collected AND system shut down. DEPRESS the BALL CATCH FLAP CATCH pushbutton on P001A (8) Turb Bldg Not MIMIC AND OBSERVE the flap rotates to the CATCH position. 113' Area 4 Required (1T203) OBSERVE ball collection starts by a rising number of balls in ball Turb Bldg 1 Not collector tank. 113' Area 4 Required (1T203) After 10 minutes STOP Ball Recirculation pump by DEPRESSING Turb Bldg .1 Not RECIRC PUMP OFF pushbutton on P001A(B) MIMIC 113' Area 4 Required (1T203) CLOSE Pump Discharge Valve F323A(B). Turb Bldg Not 113' Area 4 Required (1T203) PLACE Screens #1 AND #2 in BACKWASH position by DEPRESSING Turb Bldg 1 Not Page 24 7 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES SCREEN BACKWASH pushbutton on P001A (8) MIMIC AND 113' Area 4 Required OBSERVE screens rotate to BACKWASH position. (1T203) PRESS the CIRC WTR PMP A(B) STOP pushbutton on 1H13-P680. MCR 1 CHECK that F002A(B) Circulating Water Pump Discharge valve closes MCR on 1H 13-P680 ENSURE that A(B) Circulating Water pump has shut down USING MCR 1 pump indication light on 1H13-P680 WHEN its discharge valve is CLOSED. OPEN OR CHECK OPEN F001 USING CIRC WTR LOOP A/8 XTIE MCR handswitch on 1H13-P870. CLOSE OR CHECK CLOSED F040A (8) Acid Feed Valve. N/A N/A Not required to be performed to Shut down and Cool down the plant. CLOSE OR CHECK CLOSED LV-F513 A(B), Slowdown valve MCR OPEN F039A(B), CIRC WTR PUMP A(B) COLUMN VENT N/A N/A Not required to be performed to Shut down and Cool down the plant. ENSURE Condenser vacuum is maintained> 23.8" Hg MCR SHUTDOWN one Heater Drain Pump per SOI 04-1-01-N23-1 JOG CLOSED N23-F051A(B), HTR DR PMP A(B) DISCH VLV on 1H13-P680for desired pump. STOP HTR DR PMP A(B) on 1H13-P680. WHEN Reactor power has been reduced < 40%, SHUTDOWN 2nd Heater Drain Pmp per SOI 04-1-01-N23-1 Before securing second Heater Drain Pump, PLACE N23-LK-R053, MCR 1 HTR DR TK DR, in Manual AND Slowly REDUCE output to 0%. ENSURE Heater Drain Tank level is maintained by Dump Valves MCR N23-LV-F518A-E JOG CLOSED N23-F051 B(A) HTR DR PMP B(A) DISCH VLV on MCR 1H13-P680 for second pump. Page 248 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
- IOI / SOI ACTIONS LOCATION MODE NOTES STOP Heater Drain Pump HTR DR PMP B(A) on 1H13-P680. MCR WHEN BOTH Heater Drain Pumps are shutdown, TURB BLDG Not THEN CLOSE N23-F054, HTR DR PMP COMMON DISCH VLV on ELEV 133 Required 1H22 P175 AREA.6 ROOM 1T327.
IOI 03-1-01-2 Continued SHIFT the Reactor Recirculation Pump(s) to slow speed as follows: MCR 1 INSERT Control Rods until Load Line is between 50 AND 65% VERIFY Control Rods are in sequence of the Control Rod Pattern Controller. BEFORE entering Controlled Entry Region of Figure 3,. PERFORM the MCR 1 following WHEN TS 3.3.1.1, Action J.1 is in effect r VERIFY Fraction of Core Boiling Boundary (FCBB) is$; 1.0 per 06 RE-1J11-V-0002.
- IMPLEMENT TS 3.3.1.1, Action S2, .within 12 hours of entry AND J3 within 90 days.
IF any APRM gain is out of tolerance, THEN ADJUST gain per 06-RE- MCR 1 1C51-W-0001 prior to downshift of Recirculation Pumps. CLOSE Both Recirculation A AND B Flow Control Valves (FCV's) to MCR 1 Min Ed position using RECIRC A(B) FLO CONT on 1H13-P680 TRANSFER Both Reactor Recirculation Pumps to slow speed per MCR 1 SOI 04-1-01-833-1 CONTINUE Reactor Power reduction to 25 - 30% by insertion of MCR Control Rods SHUTDOWN Hydrogen Water Chemistry Injection per SOI 04-1-01-P73-1. At H13-P845, momentarily DEPRESS HWC SHUTDOWN pushbutton MCR 1 AND OBSERVE the following: HWC SHUTDOWN pushbutton 1P73-M602 Will be flashing as H2 AND 02 flows ramp down to 0. 02 isolation valves Will Close WHEN 02 levels remain at normal ,levels with no 02 injection for at least 5 minutes. 1 HWC SHUTDOWN pushbutton Will be in solid WHEN all control valves AND isolation valves are fully Closed. HWC RUNNING pushbutton extinguishes. CLOSE P73-F107, H2 lnj Sply Line Man Line Shutoff valve. N/A N/A Not After 02 valves F515 AND F512 (as indicated by white dots on red cap required being perpendicular to pipe) have Closed, CLOSE OR CHECK to be CLOSED Both F207 AND F208, 02 Rack Sply lsol to OG Preheater performed A(B). to Shut Page 249 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES CLOSE 1P73-F209, 02 injection to Condensate pumps down and Cool down the plant. IF Drywell entry is scheduled, WHEN Reactor Power has been reduced N/A N/A Not to less than 30%, THEN PERFORM the followin*g: required PERFORM the following for 1D21-K607, DRWL PERS HATCH ARM: to be performed to Shut DIRECT l&C to CONNECT Canon plug to the plug labeled "ALARM" down and AND "J3" at the back of 1021 K607. Cool down PLACE Function Selector switch on front of 1021 K607 (DRWL PERS the plant. HATCH ARM) to OPERATE position. PERFORM EPI 04-1-03-021-1 for 1D21K607. IOI 03-1-01-2 Continued REMOVE Both Second Stage MSR Reheaters from service per SOI 04-1-01-N11-1. OBSERVE PDS Computer Points N11 N044A,B,C AND N11 N045A,B, MCR 1 C to monitor LP Turbine Inlet Temperature OT during removal of Second Stage Reheaters from service. ENSURE Both MSR HTG STM FEED CONT are in MANUAL on MCR 1* 1H13-P680. CLOSE the following MSR 2ND STG HTG STM valves on 1H13-P680: MCR N11-F304C N11-F304D SIMULTANEOUSLY CLOSE the following MSR 2ND STG HTG STM MCR 1 valves on 1H13-P680: N11-F304A N11-F304B LOWER the manual*outputs on Both MSR HTG STM FEED CONT to MCR 1 minimum on 1H13-P680 to close the temperature control valves. CLOSE the following MSR SUPPLY VLVS valves on 1H22-P177. TURB BLDG Not N11- F028A ELEV 133 Required AREA4 N11- F028B ROOM 1T325 VERIFY the following valve lineup on local panels: N/A N/A Not N35-F015A Closed, HS-M003A required to be N35-F015B Closed, HS..:M003B performed N35-F018A Closed, HS-M007A to Shut N35-F018B Closed, HS-M007B down and Cool down the plant. Page 250 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tabl.es A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES IF Feedwater Heater 6A/B are being supplied from extraction steam TURB BLDG 1 Not (i.e., IF 1N36-F01 OA/B AND 1N36-F011A/B on 1H13-P870 are open), ELEV 133 Required* THEN CLOSE the following valves on 1H22-P177: AREA4 N35-F008A, 2ND STG RHTR DR TK A TO HTR 6A ROOM 1T325 N35-F008B, 2ND STG RHTR DR TK B TO HTR 68 REMOVE Both First Stage MSR Reheaters from service per SOI 04-1-01-N11-1. OPEN the following valves on 1H13-P870: MCR N11-F005A, MSR 1ST STG RHT RO BYP DR VLVS N11-F005B, MSR 1ST STG RHT RO BYP DR VLVS SIMULTANEOUSLY CLOSE the following valves on 1H13-P870: MCR 1 N11-F029A, HP TURB EXTR TO MSRA N11-F029B, HP TURB EXTR TO MSR B CLOSE the following valves by taking its respective handswitch to MCR 1 TEST: N11-F003A, MSR A 1ST STG RHT EXTR STM BTV (1H13-P870) N11-F003B, MSR B 1ST STG RHT EXTR STM BTV (1H13-P870) REMOVE Condensate Precoat filters from service per Not SOI 04-1-01-N22-1, IF in service. required to be performed to Shut down and Cool down the plant. OPEN* the following BSCV UPSTRM DR VLV's: MCR
- a. N33-F300A
- b. N33-F300B
- c. N33-F300C At approximately 23 - 26 % Reactor Power, RAISE the SPEED MCR DEMAND setpoint to approximately 35%, as monitored on PDS
. Computer point N32K246, by DEPRESSING the SP DEMAND RAISE AND REL pushbuttons. SIMULTANEOUSLY DEPRESS LOAD REF OFF AND REL MCR 1 pushbuttons on 1H13-P680-9C to turn off load demand Control AND VERIFY OFF light is illuminated. LOWER load by DEPRESSING SPEED DEMAND LOWER AND REL MCR 1 pushbutton. (Expected value 150-175 MWe) I OBTAIN Shift Manager permission for Manual Scram MCR 1/2 Page 251 of 270 J
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES NOTIFY the following that Main Generator is being disconnected from MCR 1/2 the grid:
- Entergy Load Dispatcher (Woodlands)
* (SMEPA) 1-601-261-2318 OR 1-601-261-2313)
- Entergy Mississippi Dispatcher )
- Duty Manager VERIFY Switchyard lineup is acceptable for trip of J5228 AND J5232 MCR 1/2 INSERT IRMs MCR 1/2 NOTIFY the following personnel/departments that a manual scram is MCR 1/2 being initiated:
- Radwaste "Chemistry
- Radiation Protection ANNOUNCE over plant pager that manual Scram is being initiated.
TAKE initial temperature data per Attachment Ill, Data Sheet I of IOI MCR 1/2 03-1-01-3 prior to scram Manually SCRAM the Reactor using the MANUAL SCRAM MCR 1/2/3 pushbuttons.
- a. VERIFY all Control Rods are fully inserted.
- b. VERIFY Reactor Power is decreasing.
- c. IF Pressure Control System is maintaining reactor pressure greater than 850 psig, THEN PLACE Reactor Mode switch to SHUTDOWN.
- d. VERIFY Reactor Recirculation pumps are running in slow speed.
ENSURE Main Turbine and Generator trip. (Reverse power 15 MCR 3 seconds time delay, 5 seconds time delay IF turbine has already tripped.).
- a. VERIFY the Generator Output Breakers open.
- b. VERIFY the Turbine Stop and Control Valves close.
WHEN reactor water level Can be restored AND maintained above 11.4 MCR 3 inches, 'THEN PERFORM the following to prevent Reactor water level from reaching Level 9 RFPT trip setpoint (58 in.): IF Reactor pressure is dropping rapidly, THEN SELECT SPEED AUTO OR MANUAL on the running Reactor Feed Pump AND LOWER Reactor Feed Pump discharge pressure to MAINTAIN Reactor level below 58 inches. TRANSFER Feedwater ConJrol to Start-Up Level Control per SOI 04-1-01-N21-1. (Attachment VII of SOI 04-1-01-N21-1 May be used.) ENSURE Scram Discharge Volume Vent AND Drain valves closed MCR 3 Page 252 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & ~-2 Bases IOI / SOI ACTIONS . IOI 03-1-01-4 SCRAM Recovery INSERT all SRM's AND VERIFY response on SRM recorders. MCR 3 SWITCH IRM/APRM LVL recorders to IRM AND VERIFY neutron MCR 3 monitoring established on IRM's IF scram signal Can be cleared AND Reactor level ANIJfpressure are MCR 3 stable, THEN RESET scram AND RETURN CRD System to normal as follows:
- BYPASS Scram Instrument Volume High Level signal by PLACING CRD DISCH VOL HI TRIP BYP switches RPS Div 1, 2, 3, 4 to BYPASS.
RESET scram by PLACING SCRAM RESET handswitches RPS Div 1, 2, 3, 4 to RESET. VERIFY all CRDs settle into Position '00'. c IF any Control Rod is NOT at the '00' position, THEN PERFORM one notch insert to attempt to force the rod to s'ettle into the '00' position. WHEN "CRD DISCH VOL WTR LVL HI TRIP" annunciator is clear, . THEN RETURN CRD DISCH VOL HI TRIP BYP switches to NORMAL. I VERIFY that the HCU scram accumulators have been recharged by OBSERVING the ACCUM FAULT indicating lights on 1H13-P680 are out. THROTILE G33-F-102 to raise bottom head drain flow AND limit MCR 3. Bottom Head Drain Line HeatupiCooldown to< 100°F/HR. Bottom head drain flow greater than 250 gpm May be required. IF Reactor water level is high, THEN REJECT water to Main Condenser per SOI 04-1-01-G33-1 to MAINTAIN level band. PLACE NSSSS OTBD MOV TEST hand§witch on 1H13-P601-19B to MCR 3 the TEST position. ...{:.. .:i*; VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 the TEST position. VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13- P601-19A-G3) Alarms. ADJUST F033, RWCU SYS BLWDN F/D CONTVLYis- 10% Open. MCR 3 OPEN OR CHE,CK OPEN the following valves: MCR 3 F028 RWCU BLWDN CTMT INBD ISOL 1H13-P680 F034, RWCU BLWDN CTMT OTBD ISOL 1H13-P680 IF rejecting to main condenser, OPEN OR CHECK OPEN in the MCR 3 following order: F046 RWCU BLWDN TO MN CNDSR 1H13-P680 Page 253 of 270
*.-.~~ Entergy
- Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown A[eas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES F041 RWCU BLWDN TO MN CNDSR BYP 1H13-P680 F235 RWCU BLWDN TO MN CNDSR 1H 13-P 870-3C F234 RWCU BLWDN TO MN CNDSR 1H13-P870-9C IF desired, while rejecting during depressurized OR low pressure MCR 3 conditions, F031, RWCU BLWDN ORF BYP VLVMay be Open to allow I' maximum flow Begin rejecting by SLOWLY OPENING F033, RWCU SYS BLWDN MCR 3 FLO CONT valve, AND IF necessary THROTTLING CLOSED F042 OBSERVE FI-R602, RWCU BLWDN FLO indicator on 1H13-P680 MCR 3 MONITOR reactor water level, blowdown flow AND area/room MCR 3 temperature indication while reject is in progress. 1 ENSURE Bypass valves are maintaining Reactor pressure MCR 3 IF proceeding to Cold Shutdown; THEN PERFORM Cooldown per MCR 3 Attachment II of IOI 03-1-01-3 concurrent with remaining steps of this attachment.
DEPRESS the MHC START DVC "LOWER" pushbutton on 1H13- MCR 3 P680-9C to reduce the MHC START DVC to Zero. CONFIRM the following Bleeder Trip valves are Closed: MCR '3
- a. N36-F013A, FW HTR SA EXTR STM BTV
- b. N36-F013B, FW HTR 58 EXTR STM BTV
- c. N36-F012A, FW HTR 6A EXTR STM BTV
- d. N36-F012B, FW HTR 68 EXTR STM BTV e: N11-F003A, MSR A 1ST STG RHT EXTR STM BTV 1
- f. N11-F003B, MSR B 1ST STG RHT EXTR STM BTV ENSURE Seal Steam Pressure AND Reactor Feed Pump operation MCR 3 maintained by main steam CLOSE the following valves as soon as possible following Turbine trip N/A N/A Not at Gas Rack 1N44D001 -N to isolate Hydrogen Pre~sure Regulators 1 required N44-PCV-F505 AND F506: to be
- a. N44-FA20 performed l to Shut
- b. N44-FA21 down and Cool down the plant.
OBSERVE the following actions occur: MCR 3 Aux Field amps AND generator output voltage indicate 0. Generator field I Primary breaker Will trip on a generator/transformer lockout condition (including Water C1rc reverse power) AND the TVR feeder switch Will open IF a lockout was - Pump can NOT initiated WHEN the Turbine speed drops to -1620 rpm. be verified running TURB AUX OIL PMPS A, B OR C starts at about 1335 rpm. by AUX PW CIRC PUMP starts at about 815 rpm. (Locally) computer Page 254 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Are~s Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES TURB SHAFT LIFT OIL PMP starts at about 510 rpm. point in TURB GEAR OIL VLVs N34-FE01/FE02 open at about 210 rpm. the MCR. THROTTLE P43-F053 to maintain Main Turbine Lube Oil temp N/A N/A Not between 90-119DF. required to be performed to Shut down and Cool down the plant. WHEN fast speed trend recording is no longer necessary AND vessel MCR 3 level is greater thary 11 .4" AND vessel pressure is less than 1064. 7 psig,, THEN PERFORM the following: DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON B21-R623A on 1H13-P601-20B. DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON B21-R623B on 1H13-P601-17B. OPEN the Generator motor operated air break GEN DISC J5230. MCR 3 PLACE Red Tag on the Control Room handswitch for J5230 in open position. (This step May be performed after step 9.30.3) AFTER GEN DISC J5230 is opened, THEN PERFORM the following: MCR 3 IF tripped, THEN RESET the following Generator reverse power relays by PRESSING the relay reset rod upwards:
- a. 432/G12 (1 N41-M752) (I
- b. 432/UT11 (1 N41-M756)
AFTER Generator reverse power relays are reset, THEN RESET the following Generator Lockout relays, IF tripped:
- a. 486-1/G12 (1 N41-M769)
- b. 486-2/G12 * (1N41-M770),
- c. 786-1/UT11 (1 N41-M759)
- d. 786-2/UT11 (1 N41-M760)
AFTER all Generator Lockout relays are reset AND IIGEN UNIT TRIP" MCR 3 annunciator clears on 1H13-P680-9A-A8, THEN OBTAIN Entergy Mississippi dispatcher's permission AND PERFORM the following to close breakers J5228 AND J5232 from 1H13-P680 panel: PLACE SYNC CONT BRKR J5228 switch to ON position. CLOSE 500 KV BRKR J5228. PLACE SYNC CONT BRKR J5228 switch to OFF position PLACE SYNC CONT BRKR J5232 switch to ON position. CLOSE 500 KV' BRKR J5232. PLACE SYNC CONT BRKR J5232 switch to OFF position. IF all Generator Lockout relays Will NOT reset, THEN PERFORM the MCR* 3 Page 255 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H~2 Bases IOI / 501 ACTIONS LOCATION MODE NOTES following: r I ,. CONTACT Electrical Maintenance to investigate reason any other Generator relays other than reverse power May have tripped. REQUEST Entergy Mississippi dispatcher to open disconnects to de-energize breaker(s) J5228 AND J5232. DEPRESS EHC SP DEMAND LOWER AND REL pushbutton on 1H13- *MCR 3 P680-9C to reduce SP DEMAND indicator to O percent. WAIT for SP LTD meter to decrease to O percent. At each Main Transformer Control Cabinet (Phase A, Phase B, AND Outside at 3 Not Phase C), MN XFMRs Required VERIFY lead cooler group fans are OFF SECURE the following steam loads to limit plant cooldown:
- SJAE per 501 04-1-01-N62-1 CLOSE Recombiner Drain Valves N64-F264 AND F265 (N64-F268 93' OG 3 Not AND F269) Pre heater Required A/B Rooms 1T1091T110 CLOSE N64-F007A(B) Preheater Inlet Drain using handswitch on N64- 113' Turb 3 Not P001. Area 1 Required 1T202 OPEN RECOMBINER AIR PURGE A(B) Manual Valve 1N64-F004A(B) 93'0G 3 Not Train A(B) Purge Air Sply Sol Byp for the corresponding recombiner Preheater Required train to ESTABLISH a purge flow of approximately 60 scfm through the A/B Rooms recombiner train. 1T109 1T110 CLOSE N62-F003A(B) CNDSR AIR TO 1 STG SJAE A(B) locally at 133' Turb 3 Not 1H22 P176 Area 1/4 Required OBSERVE that F003A(B) CNDSR AIR TO 1 STG SJAE A(B) indicates 1T305, Closed before continuing to the next step.* 1T324 DEPRESS N62-F003A(B) SJAE A(B) 1ST STG SUCT VL V CLOSE MCR 3 pushbutton on 1H13-P680 [10C].
CHECK the indication on 1H13-P680 and the following valves Close: MCR 3 SJAE A(B) 1ST STG STM INL VLV, N62-F024A(B) SJAE ICNDSR DR VLV, N62-F011A(B) SJAE A(B) 2ND STG SUCT VLV, N62-F006A(B)
\
SJAE A(B) MN STM SPLY VLV, N62-F001A(B) SJAE A(B) EXH VLV, N62-F012A(B) SJAE A(B) SEP DR VLV, N62-F002A(B) REDUCE setpoint of N62-PIC-R01 OA(B) to zero O psi 113' Turb 3 Not Area 1 Required 1T202 Page 256 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES ENSURE OPEN following handswitches on 1H22-P176: 133' Turb 3 Not N62-F004A the COND AIR TO 1 STG SJAE A Area 1/4 Required N62-F004B, the COND AIR TO 1 STG SJAE B 1T305, 1T324 ENSURE OPEN N62-F034 A, B, C, DISCH PIPE ORN VLV for draining 1J3' Turb A 3 Not discharge piping. MVP Area Required 1T218 WHEN discharge piping has drained, 113' Turb A 3 Not THEN CLOSE N62-F034 A, B, C, DISCH PIPE ORN VLV. MVP Area Required U218 OPEN N62-F014 MECHVAC PUMPS COM SUCTVLV, at 1H22-P176. 133' Turb 3 Not Area 1/4 Required 1T305,
" 1T324 ENSURE proper mechanical vacuum pump oil *level (>50%), 113' Turb A 3 Not THEN Prelube with manual oiler as follows: MVP Area Required ENGAGE manual oiler pump handle 1T218 ROTATE for a minimum of 60 seconds.
DEPRESS each plunger 5 times CHECK oil flow visible from each oil return line. CLOSE P41-F348 A(B,C) MECH VAC PMP COOLER DRAIN. 113' Turb A 3 Not OPEN P44-F109 A(B,C) MECH VAC PMP PSW INL ISOL. MVP Area Required I OPEN P44-F344 A(B,C) MECH VAC PMP PSW DISCH ISOL. 1T218 BLOW DOWN strainer as follows: (1) OPEN P4;4-F316 A(B,C), MECH VAC PMPA(B)(C) STR DR. (2) WHEN blowdown has been completed, THEN CLOSE P44-F316 A(B,C) MECH VAC PMPA(B)(C) STR DR. START MECH VAC PMP A(B)(C) with START pushbutton on 1H13 MCR 3 P680. CHECK proper vacuum pump operation for each running pump by 113' Turb A 3 Not OBSERVING the following: MVP Area Required Cooling Water Inlet Valve 1P44-SV-F514A, B, OR C has opened by 1T218 MOMENTARILY OPENING drain valve 1P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR. OBSERVING pressurized water flow, THEN CLOSE drain valve1 P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR. IF 1P44-SV-F514A, B, OR C did NOT open, . THEN OPEN respective MECH VAC PMP A(B)(C) PSW SPLY BYP valve 1P44-F347A,B,C to provide cooling as needed for operation of Mechanical Vacuum Pump. Suction Drain Valve SV-F507A, .B OR Chas Closed by OBSERVING Page 257 of 270 ,
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES no air suction flow.
Mechanical Vacuum Pump Inlet Valve F007A, B, OR Chas Opened. Proper oiler operation by OBSERVING oil flow from each oil return line. Secure Seal Steam Generator per SOI 04-1-01-N33-1 PLACE Controller PK-R617 in MANUAL on 1H13-P878, AND CLOSE F506 as necessary to control reactor cooldown. The turbine Can be sealed with seal steam header pressure as low as 15 psig, PI-R622. Secure Reactor Feed Pump per SOI 04-1-01-N21-1 IOI 03-1-01-4 Continued Off gas Preheater by placing controllers 1N64-R009A and 1N64-R009B Turbine 3 Not in manual and reducing output to O percent Building 93' Required Area 1 (1T113) Main Steam Isolation valves AND/OR Main Steam Line MCR 3 SHUTDOWN a Condensate Booster Pump AND CLOSE respective MCR 3 discharge valve per SOI 04-1-01-N19-1, leaving one Condensate Booster Pump in service SHUTDOWN a Condensate Pump AND CLOSE respective discharge MCR 3 valve per SOI 04-1-01-N19-1, leaving one Condensate Pump in service CLOSE B21-F069 MCR 3 OPEN the following MSIV drain valves:
- a. B21-F067A
- b. B21-F067B
- c. B21-F067C
- d. B21-F067D OPEN the following valves: MCR 3
- a. B21-F033
- b. B21-F068 ISOLATE extraction steam to the HP Feedwater heaters as follows: MCR 3 CLOSE: the following valves:
- a. N36;-F01 OA EXTR STM SPLY TO FW HTR 5A
- b. N36-F01 OB EXTR STM SPLY TO FW HTR 58
- c. N36-F011A EXTR STM SPLY TO FW HTR 6A
- d. N36-F011B EXTR STM SPLY TO FW HTR 68 Page 258 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MOD_E NOTES I OBSERVE the following drain valves open: MCR 3 N36-F008A\ FW HTR 6A EXTR STM RO BYP DR VLV N36-F0088 FW HTR 68 EXTR STM RO BYP DR VLV OPEN the following drain valves: MCR 3 OPEN HP Stop AND Control Valve Drain Valves by DEPRESSING each of the followingMSCV UPSTRM DR VLV "JOG OPEN" pushbuttons:
- a. N33-F078A
- b. N33-F0788
- c. N33-F078C
- d. N33-F078D OPEN Left Side Crossover piping drains by DEPRESSING each of the following XOVER PIPE LS DR VLV "J09 OPEN" pushbuttons:
- a. N11-F043A (FRIST) ',
- b. N11-F036A (FR 2ST)
- c. N11-F044A (RE IST)
- d. N11-F038A (RE2ST)
OPEN Right Side Crossover piping drains by DEPRESSING each of the following XOVER PIPE RS DR VLV "JOG OPEN" pushbuttons:
- a. N11-F0448 (FRIST)
- b. N11-F0388 (FR 2ST)
. c. N11-F0438 (RE IST)
- d. N11-F0368 _(RE 2ST)
OPEN N11-F015, MSCV A/8 DNSTRM DR VLV. OPEN the following MSR 2ND STG STM DR VLVS:
- a. N11-F301
- b. N11-F302 OPEN the following drain valves unless required closed to minimize MCR 3 cool down:
OPEN Main Steam Line Drain Valves N11-F056, F055, F009, F011, F049, AND F050 by DEPRESSING MSL DR LINE IS'OL VLVS "OPEN" pushbutton. OPEN MSL Bypass Drain valves (N11-F002A, F0028, F002C, F002D, .F010, FOO?, F052A, F0528, F057) using MSL DR VLVS DR LINE BYP VLV "OPEN" pushbutton. DEPRESS Both NSSSS INBD ISOL RESET pushbutton (1H13-P601- MCR 3 188) AND NSSSS OTBD ISOL RESET pushbutton (1H13-P601-198) to reset logic AND re-energize RHR Logic lights on 1H13-P622 AND 1H13-P623 panels. TRANSFER to startup level control IF NOT already in service MCR 3 TRANSFER the RFPT A(B) SP CONT to MAN. MCR 3 Page 259 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases ( IOI / 501 ACTIONS LOCATION MODE NOTES IF MSIV's are open with Main Condenser available, THEN INITIATE MCR 3 AND MAINTAIN cooldown at:;:;; 90°F/hr with one of the following methods: CONTROL Reactor cool down with Manual Bypass Jack 'on 1H13-P680-9C At approximately 200 psig Reactor pressure, SHUTDOWN one RWCU Pump per SOI 04-1-01-G33-1, IF Both are running. SLOWLY OPEN 1G33-F044, RWCU FLTR DMIN BYP VLVon 1H13- MCR 3 Not P680 while reducing F/D flow with flow controller 1G36-FC-R022A(B) CTMT 185' Required on 1G36-P002. RWCU Panel (1A509) MAINTAIN a nearly constant system flow rate, (450-500 gpm MCR 3 Is recommended), as indicated on 1G33-FI-R609, RWCU INL FLO, on1H13-P680. On 1G36-P002, OBSERVE that holding pump comes on WHEN F/0 CTMT 185' 3 Not flow is< 80%. RWCU Panel Required (1A509) WHEN filter flow is < 20%, TURN Filter/Hold switch A(B) on 1G36-P002 CTMT 185' 3 Not to HOLD position. RWCU Panel Required OBSERVE the following valves fully Close: (1A509) G36-F001A(B) F/0 Inlet G36-F002A(B) F/0 Inlet G36-F003A(B) F/D Outlet G36-F004A(B) F/0 Outlet OBSERVE HOLD light on AND Fl LTER light out on 1G36-P002 CTMT 185' 3 Not RWCU Panel Required (1A509) PLACE the MANUAUAUTO selector on controller 1G36-FC-R022A (B) CTMT 185' 3 Not in MANUAL position with controller output at 0% output. RWCU Panel Required (1A509) REPEAT Steps 4.6.2a AND 4.6.2b for second F/D. CTMT 185' 3 Not RWCU Panel Required (1A509) LOWER system flow rate to < 280 gpni by THRO TILi NG 1G33F044 as MCR 3 indicated on 1G33FI-R609, RWCU INL FLO, on 1H13-P680. TRIP one of the running RWCU pumps MCR 3 ESTABLISH 90 to 300 gpm flow as indicated on 1G33-FI-R609, MCR 3 RWCU INL FLO, on 1H13-P680 by THROTILING the Bypass Valve 1G33F044 WHEN Reactor pressure is reduced to< 135 psig, THEN at Page 260 of 270
. ~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 ~ Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / 501 ACTIONS
- approximately 40 psig, PLACE one loop of RHR System in SHUTDOWN COOLING mode per 50104-1-01-E12-~.
RACK OUT RHR A/B PMP Breaker, 152-1509/1606 Control Bldg. 3 Required 111' SWGR Rms OC202, OC215 SHUTDOWN RHR JOCKEY PUMP A/Bon 1H13-P871. MCR 3 CLOSE F082A/B, RHR JCKY Pl\i1P SUCT ISOL VLV, on 1H13-P871. MCR 3 CLOSE F064A/B, RHR MIN FLO TO SUPP POOL. MCR- 3 CLOSE F004A/B, RHR PMP SUCT FM SUPP MCg 1* 3 ENSURE OPEN F003A/B, RHR HX QUTL VLV. MCR 3 ENSURE OPEN F048A/B, RHR HX A BYP VLV. MCR 3 CLOSE F047A/B, RHR HX INL VLV. MCR 3 CLOSE F428A/B, PRESSURE LOCK ISOL for F024 RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) CLOSE F438A/B; PRESSURE LOCK ISOL for F064 RHRA/B 3 Required.
. /
Pump Rm Aux Bldg 93' (1A 103/1A10 5) SLOWLY OPEN F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC Rm (1A204) OPEN F006A, RHR PMP A SUCT FM SHUTDN CLG AND MONITOR MCR 3 RHR HR A STM press indicator for rise in pressure. VENT Shutdown Cooling suction header as follows: Aux Bldg 3 Required (a) OPEN F323. 119' RCIC Rm (1A204) (b) OPEN F399. (c) WHEN a solid stream of water is observed out of vent line, THEN CLOSE F399. (d) CLOSE F323. OPEN F073A, RHR HX A OTBD VENT VLV. MCR 3 OPEN F074A, RHR HX A INBD VENT VLV. MCR 3 VENT RHR A Heat Exchanger A as follows: Aux. 139' 3 Required (a) OPEN F400A, A RHR HX VENT. RHRA/B Rm Page 261 of 270
--=====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES (b) OPEN F401A, A RHR HX VENT. 1A303, (c) WHEN water is observed from vent, THEN CLOSE F401A. 1A304/1A306
, 1A307 (d) CLOSE F400A.
OPEN F064A/B. AFTER approximately one minute, THEN CLOSE MCR 3 F064A/B WHEN Conductivity as indicated on HX A/BOUT CNDCT, is as low as MCR 3 practical (Should be less than 2.0 µmhos/cm), THEN CLOSE F073A/B, RHR HX A/B OTBD VENT VLV. CLOSE F074A/B, RHR HXA/B INBD VENT VLV MCR 3 LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC
'\
Rm (1A204) CLOSE F048A/B MCR 3 OPEN F063A/B, Manual Flush Valve. RHRA/B 3 Required Pump Rm Aux Bldg 119' (1A203/1A20 5) OPEN F073A/B, RHR HX A/B OTBD VENT VLV MCR 3 OPEN F074A/B, RHR HXA/B INBD VENTVLV MCR 3 WHEN Conductivity as indicated on HX A/B OUT CNDCT, is as low as MCR 3 practical (Should be less than 2.0 µmhos/cm), THEN CLOSE F073A/B, RHR HX A/B OTBD VENT VLV. CLOSE F074A/B, RHR HXA/B INBD VENT VLV MCR 3 LOCK CLOSED F063A/B, Manual Flush Valve. RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A203/1A20 5) OPEN F048A. MCR 3 OPEN F047A. MCR 3 ENSURE OPEN F003A. MCR 3 ENSURE Shutdown Cooling Isolation Logic is reset by PRESSING MCR 3 NSSSS INBD ISOL RESET pushbutton AND NSSSS OTBD ISOL RESET pushbutton on 1H13-P601. PLACE Standby Service Water A System in service to RHR A Heat MCR 3 Exchanger on 1H13-P870 as follows. START SSW Pump A per SOI 04-1-01-P41-1. Page 262 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
>t-- *.:.a,.:{:;;~r~!(iiiJnent 2- Safe Operation & Shutdown Areas Tables A-~ & H-2 Bases
,~. . . ** . :. :/'~*). ?~: ,yJ .f:*.-1:'.~r*.-
IOI / SOI ACTIONS LOCATION MODE NOTES I OPEN P41-F014A, SSW'INL TO RHR HXA. ENSURE OPEN P41-F068A, SSW OUTL FM RHR HX A. START RHR RM A FAN COIL UNIT. ENSURE OPEN F010, SHUTDN CLG MAN SUCT VLV. MCR 3 ENSURE CLOSED F040, RHRTO RADWST OTBD SHUTOFF VLV. MCR 3 ENSURE CLOSED F049, RHR TO RADWST INBD SHUTOFF VLV. MCR 3 OPEN F020, Manual Flush Valve approximately 3 turns. Valve May be Aux Bldg 3 Required opened further IF required.for level control. 119' RCIC Rm (1A204) OPEN F008, RHR SHUTDN CLG OTBD SUCT VLV MCR 3 OPEN F009, RHR SHUTDN CLG INBD SUCT VLVas follows; MCR 3
.ENSURE breaker 52-163137 is CLOSE: position 1 OPEN F009, RHR SHUTDN CLG INBD SUCT_VLV MONITOR Reactor water level WHILE 1E12F009AND 1E12F020 are OPEN.
PERFORM IMMEDIATELY the next step 4.1.2.b(14) IF a rise in Reactor water level is NOT desired. LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC Rm (1A204) I NOTIFY Radwaste Operators to be prepared for Reactor water flush to MCR 3 Required Waste Surge tank. Radwaste Building 118' Radwaste Control Room (OR241) OPEN F203, RHR SYS FLUSH TO LIQ ~DWST by the following MCR 3 handswitches to OPEN: F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1 H13-P870-3C) F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C) ENSURE CLOSED F070A/B, Manual RHR Drain Valve Aux Bldg 93' 3 Required Corridor (1A101) -* OPEN F072A/B, RHR Drain Valve RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A103/1A10 ( 5) Page 263 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tabfes A-3 & H-2 Bases IOI / 501 ACTIONS LOCATION MODE NOTES SLOWLY OPEN F070A, RHR Drain Valve approximately one turn to Aux Bldg 93' 3 Required start flow to Radwaste. Corridor IF "RHR A DISCH PRESS ABNORMAL" annunciator alarms while (1A101) warming RHR A, THEN CLOSE F047A AND F048A to prevent draining of downstream piping.
THROTTLE F070A/B to warm RHR Pump A/B at less than 100°F/hr Aux Bldg 93' 3 Required until RHR DISCH TO RADWST ON RHR TEMP recorder is 200°F OR Corridor within 100°F of RX water temp, whichever is less. (1A101) LOCK CLOSED F070A, RHR Drain Valve. Aux Bldg 93' 3 Required Corridor (1A101) LOCK CLOSED F072A, RHR Drain Valve. RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) CLOSE F203, RHR SYS FLUSH TO LIQ RADWST by TAKI NG the MGR 3 following handswitches to CLOSE: F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1 H13-P870-3C) F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C) RACK IN RHR A/B PMP Breaker, 152-1509/1606 Control Bldg. 3 Required 111' SWGR Rms OC202, OC215 NOTIFY Chemistry AND Radiation Protection that possibility of a MGR 3 crud burst Could occur due to starting of RHR pump in SOC mode START OR ENSURE running RHR RM A FAN COIL UNIT on 1H13- MGR 3 P870. ENSURE CLOSED F064A, RHR A MIN FLO TO SUPP POOL. MGR 3 ENSURE RHR JOCKEY PUMP A is shutdown. MGR 3 ENSURE CLOSED F082A, RHR A JCKY PMP SUCT ISOL VLV. MGR 3 ENSURE CLOSED F004A, RHR A SUCT FM SUPP POOL. MGR 3 ENSURE OPEN the following valves: MGR 3 (a) F010 (Concurrent Verification Required) (b) FOOS (c) F009 as follows; (1) ENSURE breaker 52-163137 is CLOSE position (2) ENSURE OPEN F009, RHR SHUTDN CLG SUCT VLV (d) F006A Page 264 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
- Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES (e) F047A (f) F048A I CLOSE F003A, RHR HX A OUTL VLV. MCR 3 ENSURE CLOSED B21-F065A, FW INL SHUTOFF VLV. MCR 3 START RHR l?MR A AND IMMEDIATELY FULLY OPEN, one of MCR 3 the following valves:
(a) E12-F053A, RHR A SHUTDN CLNG RTN TO FW (b) E12-F037A, RHR A TO CTMT POOL (c) 'E12-:F042A, RHR A INJ SHUTOFF VLV , MONITOR RHR HX A differential temperature on RHR MCR 3 TEMPERATURE RECORDER as follows:
'RHR HXA Point 1(inlet) - Point 5(outlet)
ESTABLISH a cool down rate of less than 90°F/hr, as follows: MCR 3 Slowly JOG OPEN F003A to allow flow through heat exchanger, AND MONITOR cooldown rate. THROTILE one of.the following valves to maintain RHR pump flow
-8600 gpm AND RHR heat exchanger flow -8200 gpm:
IF flow is through F053A, THEN THROTILE F053A AS LONG AS flow through valve is maintained < 8550 gpm. IF E12-F003A is closed while in SHUTDOWN COOLING, MCR 3 THEN MONITOR REAyTOR COOLANT TEMPERATURE using the following indications: REACTOR RECIRC LOOP A/B suction temperature (IF recirc pump(s) running) RWCU REGENERATIVE HEAT EXCHANGER INLET temperature (IF RWCU pump(s) are running.) Point 5 qf RHR TEMPERATURE RECORDER. Installed thermocouple suspended above Reactor core. WHEN F003A valve is full open AND additional cooling is required, MCR 3 THEN SLOWLY THROTILE CLOSE F048A as needed to establish desired cooldown rate. WHEN F048A valve is full closed, THEN, IF desired, THROTILE MCR 3 F003A to MAINTAIN desired coolant temperature OR SOC flow while MAINTAINING~ 3000 gpm flow. F048A may be fully opened to reduce cooldown rate but CANNOT be left in a throtUed position UNTIL F003A is full open: SELECT "Shutdown Cooling-RHR A" OP GUIDE on PDS computer. MCR 3 The guide Should be left on-screen OR icon'd WHEN the respective ' shutdown cooling loop is in service until Reactor Coolant has been stabilized at desired temperature so that the guide Will warn operators IF Shutdown Cooling parameters are out of range I Page 265 of 270
.~Entergy
- Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / 501 ACTIONS LOCATION MODE NOTES LOG Reactor coolant temperature on Data Sheet I of 03-1-01-3 OR MCR 3 other applicable IOI. TAKE temperatures as required by 03-1-01-3 during cooldown AND CONTINUE to take readings once per hour W~EN temperature is stable.
LOG temperatures for SSW/RHR HX AND reactor coolant on log MCR" 3 similar to Attachment I to ENSURE SSW temperature does NOT exceed design temperatures. (Ref. CR1997-0282) IF SSW A auto start signal from RHR A pump running is defeated by Temporary Alteration, THEN START/STOP SSW A AND B fans as necessary to MAINTAIN SSW A Supply temp. (E12-R601, pt. 12) between 50 AND 75 deg. IF RPV level control via RWCU blowdown is unavailable, MCR 3 THEN RPV level control May be established by USING E12-F073A AND E12-F074A RHR heat exchanger vent to establish RPV level control, AND THROTTLE OPEN E12-F073A AND E12-F074A as required to establish AND maintain the desired RPV level. MONITOR RPV level while reject is in progress. IF desired to add water to Reactor with SDC in operation WHEN in 1 Aux Bldg 4, 5 Not Modes 4 OR 5, THEN PERFORM the following: 119' RCIC Required THROTTLE OPEN, F020. Rm (1A204) WHEN desired Reactor Vessel Level is reached, THEN LOCK CLOSED F020. IOI 03-1-01-3 Continued At approximately 120 psig, PERFORM the following: TRANSFER RWCU to Pre-pump mode per 501 04-1-01~G33-1. PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601-19B to MCR 3 the TEST position. VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 the TEST position. VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13-P601-19A:-G3) Alarms. SECURE RWCU blowdown flow per Section 5.1 of this instruction. MCR 3 STOP running RWCU pump AND leave F044, RWCU FLTR DMIN BYP MCR 3 VLV THROTTLED SLIGHTLY OPEN. CLOSE the following valves AND proceed to Step 4.4.2g without delay: MCR 3 F250 RWCU SPLY TO RWCU HXS 1H13-P870-3C F251 RWCU SPLY TO RWCU HXS 1H 13-P870-9C Page 266 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES F252 RWCU HX RTN TO RWCU PMPS 1H13-P870-9C F253 RWCU HX RTN TO RWCU PMPS 1H13-P870-3C F255 RWCU FLTR/DMIN INL FM RWCU PMP 1H13-P870-5C OPEN OR CHECK OPEN the following valves: MCR 3 F004 PMP SUCT CTMT OTBD ISOL 1H13-P680 F001 PMP SUCT DRWL INBD ISOL 1H13-P680 F254 RWCU FLTR/DMIN INL FM RWCU HX 1H13-P870-5C F256 RWCU HX INL FM RWCU PMP 1H13-P870-5C *. .* : . CLOSE OR CHECK CLOSED F044, RWCU FLTR DMIN BYP VLV; MCR 3 And THEN RESTART one RWCU pump AND JOG OPEN F044 to establish flow greater than 90 gpm but less than 300 gpm. START one RWCU the pump AND THROTTLE F044 to achieve a MCR 3 system flow greater than 90 gpm, But less than 300 gpm as indicated on FI-R609; RWCU INL FLO: IF performing system warm-up. THEN MAINTAIN minimum flow, AVOIDING low flow trip. IF desired, START a second pump as follows: MCR 3 START the pump AND THROTTLE F044 to maintain 300 - 500 gpm system flow as indicated on FI-R609, RWCU INL FLO, with Both Pumps running. IF desired, ESTABLISH RVVCU blowdown flow in accordance with All areas Section of this instruction previously addressed for this evolution IF desired, PLACE F/Ds in service in accordance with Section 4.5 of All areas this instruction. previously addressed for this evolution PLACE NSSSS OTBD MOV TEST handswitch on 1H 13-P601-19B to MCR 3 the NORM position. VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Clears. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 '" the NORM position. VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13- P601-19A-G3) Clears. IOI 03-1-01-3 Continued SHUTDOWN the running Condensate Booster Pump AND CLOSE respective discharge valve per SOI 04-1-01-N19-1. Page 267 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS
,;* ~
NOTES IF scheduled, THEN PERFORM 06-0P-1821-R-0010 (Att. I AND/OR II) Not WHEN reactor pressure is between 50 AND 100 psig required to be performed to Shut down and Cool down the plant. At approximately 60 psig Reactor pressure, PERFORM the following: MGR 3 VERIFY that RCIC system isolates automatically. IMMEDIATELY NOTIFY GAS, SAS, OR Security Island that RCIC is not available (non-functional). COMPLETE shutdown of RCIC system per SOI 04-1-01-E51-1. WHEN cooldown using Bypass Valves is no longer desired AND MGR 3 Shutdown Cooling is in service, THEN CLOSE the Bypass Valves as follows: SET the TURB STM PRESS DEMAND setpoint approximately 100 psig above Reactor pressure using the PRESS REF "RAISE" OR "LOWER" pushbuttons on 1H13-P680-9C. DEENERGIZE the Manual Bypass Valve Controller by depressing the MAN BYP CONT "OFF" pushbutton on 1H13-P680-9C. IF MSIV's are open AND stroke time testing was NOT scheduled, MGR 3 THEN PERFORM the following: CLOSE the following Inboard MSIVs: B21-F022A B21-F022B B21-F022C B21-F022D WHEN Main Steam Line pressure downstream of MSIVs is near zero psig, THEN CLOSE the following Outboard MSIVs: B21-F028A B21-F028B B21-F028C B21-F028D CLOSE B21-F016 . CLOSE B21-F019 NOTIFY Radiation Protection that the Reactor is to be vented to *McR 3 Drywell sump AND REQUEST Drywell survey after Head Vent realignment. WHEN Reactor coolant temperature is less than 210°F, THEN MGR 3 REALIGN Reactor Head Vents on 1H13-P601 as follows: OPEN 1B21-F001, RPV OTBD VENTVLV. Page 268 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE** NOTES PEN 1 B21-F002, RPV INBD VENT VLV. CLOSE 1821-FOOS, RPV VENT TO MSL A. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2. Page 269 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & M-2 Bases Table A-3 & H-2 Results Table A-3 & H-2 Safe Operation & Shutdown Rooms/Areas
- Room/Area Mode Control Building 111' SWGR Rms (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A 103) 3 Auxiliary Building 93' RHR B Pump Room (1A105) 3 Auxiliary Building 93' Corridor (1A 101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203) 3 Auxiliary Building 119' RHR B Pump Room (1A205) 3 Auxiliary Building 119' RCIC Room (1A204) 3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR B Room (1A306, 1A307) 3 Radwaste Building 118' Radwaste Control Room (OR241) 3 Page 270 of 270
GNR0-2018/00048 Page 19 of 19 GNR0-2018/00048, E~CLOSURE ATTACHMENT 3 GGNS EAL BASIS DOCUMENT MARKUP
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Grand Gulf Nuclear Station EAL Technical Basis Page 1 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Table of Contents I
1.0 INTRODUCTION
- 3 2.0 DISCUSSION ............................................................................................................. 3 2.1 Background ...................................................................................................... 3 2.2 Fission Product Barriers ................................................................................... 4 2.3 Fission Product Barrier Classification Criteria ................................................... 4 2.4 EAL Organization .................................................................... :........................ 5 2.5 Technical Bases Information ............................................................................ 7 2.6 Operating Mode Applicability ............................................................................ 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATION S .................................. 9 3.1 General Considerations .................................................................................... 9 3.2 Classification Methodology ............................................................................. 10
4.0 REFERENCES
........................................................................................................ 14 4.1 Developmental ................................................................................................ 14 4.2 Implementing." ...................................................................................... :'.......... 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS .................................................. 15 5.1 Definitions (ref. 4.1.1 except as noted) ........................................................... 15 5.2 Abbreviations/Ac ronyms ................................................ ;............................ ~2-0 6.0 GGNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERE NCE ................................... 24~
7.0 ATTACHMENTS ...................................................................................................... 27 7.1 Attachment 1, Emergency Action Level Technical Bases ............................... 28 Category A-Abnormal Rad Levels I Rad Effluent ....................................... 28 Category C - Cold Shutdown I Refueling System Malfunction ...................... 65 Category E - Independent Spent Fuel Storage Installation (ISFSI) ...... 1064G5 Category F - Fission Product Barrier Degradation ................................ 109400 Table F-1 Fission Product Barrier Threshold Matrix & Bases 113 Category H - Hazards and Other Conditions Affecting Plant Safety ..... 1684@.7 Category S - System Malfunction ...................................................~ ........... 203 7.2 Attachment 2; Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases 243 Page 2 of 270
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-::::::- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
1.0 INTRODUCTION
This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Grand Gulf Nuclear Station (GGNS). It should be used to facilitate review of the GGNS EALs and provide historical documentation for future reference. Decision-makers respons.ible for implementation of 10-S-01-1, Activation of the . Emergency Plan, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification. Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRG staff expects th~t changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the GGNS Emergency Plan. In 1992, the NRG endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: *
- Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
- Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).
- Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRG EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Actior, Levels Page 3 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX for Non-Passive Reactors," November 2012 (ref. 4.1.1 ), GGNS conducted an EAL implementation upgrade project that produced the EALs discussed herein. 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containme_nt of radioactive materials. A "Potential Loss" threshold implies a greater probability of barrier loss and reduced certainty of maintaining the barrier. The primary fission product barriers are: J A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to
,and including the isolation valves.
- C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are _used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency.
2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert:
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Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier Page 4 of 270
-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 2.4 EAL Organization The GGNS EAL scheme includes the following features:
- Division of the EAL set into three broad groups:
o EALs applicable .under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode. o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
- Within each group, assignment of EALs to categories' and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The GGNS EAL categories are aligned to and represent the NEI 99-01 "Recognition Categories." Subcategories are used in the GGNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The GGNS EAL categories and subcategories are listed below. The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information. Page 5 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX EAL Groups, Categories and Subcategories EAL Group/Category EAL ~ubcategory I Any Operating Mode: A -Abnormal Rad Levels I Rad Effluent 1 - Radiological Effluent
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2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1- Security Affecting Plant Safety 2- Seismic Event 3- Natural or Technological Hazard 4- Fire 5- Hazardous Gas 6- Control Room Evacuation 7- Emergency Director Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI) Hot Conditions: S - System Malfunction 1- Loss of ESF AC Power 2- Loss of Vital DC Power 3- Loss of Control Room Indications 4- RCS Activity 5- RCS Leakage 6- RPS Failure 7- Loss of Communications 8- Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown I Refueling System 1 -RPV Level Malfunction 2 - Loss of ESF AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems Page 6 of 270
~Entergy . GrandL Gulf N*uclear Station EAL Basis Document Revision 1 xxx 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, Hand S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 ~ev. 6.. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
- 1. First character (letter): Corresponds to the EAL category as described above (A, C, E, F, Hor S)
- 2. Second character (letter): The emergency classification (G, S, A or U)
G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event
- 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, this character is assigned the number one (1 ).
- 4. Fourth character (number): The numerical sequence of the EAL within the EAL*
subcategory-. If the subcategory has only one EAL, it is given the' number one* (1 ). Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix.
. (
Page 7 of 270
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~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled, or All. (S~e Section 2.6 for operating mode definitions)
Definitions: If the EAL wording contains a defined term, the definition of the term is included in this
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section. These definitions can also be found in Section 5.1. Basis: An EA'L basis section that provides GGNS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Reference(s): Source documentation from which the EAL is derived 2.6 Operating Mode Applicability 1 Power Operation Reactor is critical and th~ mode switch is in RUN 2 Startup The mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN and average reactor .coolant temperature is
>200°F 4 Cold Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is
~ 200°F 5 Refueling The mode switch is in REFUEL or SHUTDOWN with one or more reactor vessel head closure ,bolts are less than fully tensioned DEF Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
Page 8 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs ar~ based on loss or potential loss of Fission Product Barrier Thresholds. EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining thisI verification would exceed the 15 minute declaration , requirement. Place keeping should be used on ,all EAL matrices. 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare-an em~rgency condition within 15 minutes,after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8). 3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the. indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for ttmely assessment. 3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time dura_tion (e.g., 15 minutes, 30 minutes, etc.), the* Emergency Director should not wait until the applicable time has. elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed; the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. Page 9 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX r
3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant .an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or . component. In these cases, the controls associated with the planning, preparation .and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50. 72 (ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling*, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Emergency Director Judgment I While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H)( The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Class.ification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If am EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the EC.L must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8). Page 10 of 270
--===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: ** If an Alert EAL ahd a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
- If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.
- 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not relat.ed to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot*Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency cla$sification should be made as if the 1 EAL has been met. While applicable to all emergency classification levels, this approach is. particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. Page *11 of 270
-===- .Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 3.2.4 Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met.
3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of'declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram. 3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration,*then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision Page 12 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2. 7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applic~ble. Specifically, the event should be reported to the NRG in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRG is discussed in NUREG-1022 (ref. 4.1.3). Page 13 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
4.0 REFERENCES
4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50. 72 Immediate Notification Require.ments for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50. 73 License Event Report System 4.1.6 GGNS Technical Specifications Table 1.1-1, Modes ( 4.1. 7 GGNS Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 GGNS Emergency Plan 4.1.10 GGNS UFSAR 9.1.4.2.10.4 Storage of Fuel at the Independent Sp~nt Fuel Storage Installation (ISFSI) 4.1.11 GGNS UFSAR 9.1.4.2.1 O Description of Fuel Transfer 4.1.12 SOPP-018-1 Shutdown Operations Protection Plan 4.1.13 1O-S-01-12 Radiological Assessment and Protective Action Recommendations 4.2 l,rnplementing 4.2.1 10-S-01-1 Activation of the Emergency Plan 4.2.2 NEI 99-01 'Rev. 6 to GGNS EAL Comparison Matrix 4.2.3 GGNS EAL Matrix Page 14 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX, 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted) Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital lette(s (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or ar security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC) (ref. 4.1.10). Containment Closure The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under, shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established (ref. 4.1.12). Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
- Unusual Event (UE)
- Alert
- Site Area Emergency (SAE)
- General Emergency (GE)
Page 15 of 270 '
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~~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.
Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. \ Flooding A condition where water is entering a room or area faster than installed* equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the I SECURITY OWNER CONTROLLED AREA). Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
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Grand Gulf Nuclear Station EAL Basis Document Revision XXX Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. , lmpede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating Condition (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. o,,,~mer Controlled Area (OCA) For the purposes of classification, the Security area berNeen the QC/\ detection fence and the PROTECTED AREA boundary known as the Security Ovmer Controlled Area (SOGA) in the GGNS Emergency Plan (ref. 4 .1.9). Projectile An object directed toward a Nuclear Power Plant thatcould cause concern for its continued operability, reliability, or personnel safety. Protected Area An area encompassed by physical barriers (i.e., the security fence) and to which access is
- controlled. (ref. 4.1.9).
RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway 1 Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway (ref. 4.1.11). Restore Take the appropriate action required to return the value of an 'identified parameter to the applicable limits. 1 Page 17 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFRS0.2):
Those structures, systems and components, that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threaUrisk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION. Security Owner Controlled Area (SOCA) The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY. Site Boundary That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor (ref. 4.1.13) Unisolable An open: or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event
/
Page 18 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Events are in progress or have occurred which indicate a potential. degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material .requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs..
Page 19 of 270
~Entergy r*
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. I Visible Damage* Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Page 20 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 5.2 Abbreviations/Acronyms
~F ....................................................................................................... Degrees Fahrenheit 0
************************************************************************************************************:************** Degrees AB ............................................................................................................Auxiliary Building AC ........................................................................................................ Alternating Current AOP ................................................................................. Abnormal Operating Procedure APRM .................................................................................. Average Power Range Meter ARI ....................................................................................*............ Alternate Rod Insertion A TWS ...................................................................... Anticipated Transient Without Scram BWR ............................................................................................... Boiling Water Reactor BWROG ......................... :........................................ Boiling Water Reactor Owners Group CDE .............................. ;........................................................ Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CNB ............................................................ ~.************************************** Containment Barrier CS .................................................................................................................... Core Spray CTMT .............................................. ~ ........ ,..................................................... Contai*nment DEF .....................................................................................................................Defueled OBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current DIG ......................................................................................................... Diesel Generator EAL ............................................................................................. Emergenqy Action Level ECCS ............................................................................ Emergency Core Cooling System ECL. ............... *.* ............................... :................................ Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPG ................................................................................ _Emergency Procedure Guideline EPP*********************************************************************************,******* Emergency Plan Procedure
- ERO ................................. ~ ......................................... Emergency Response Organization ESF ................,. ......................................*.................................. Engineered Safety Feature FAA .................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FCB ........... :............................................................................................ Fuel Clad Barrier FEMA. ............................................ :........... : ..... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE .... .\ ................................................"............................................... General Emergency Page 21 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
_j HCTL ............................................................................ Heat Capacity Temperature Limit HPCS ............. ***********************************************:************************** High Pressure Core Spray IC ...............................................................................................:.......... lnitiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI. ........................................................... Independent Spent Fuel Storage Installation Kett ..................... ..................... ..................... .......... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER. ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPCS ........................................................................................ Low Pressure Core Spray LRW.............................. :***************************************: .................................. Liquid Radwaste LWR. .................................................................................................. Light Water Reactor MPC ................................... Maximum Permissible Concentration/Multi-Purpose Canister MPH ...................................................................................................-: ........ Miles Per Hour mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man M_SCRWL. ....................................................... Minimum Steam Cooling RPVWater Level MSIV ....................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line MW .......................... ~ ......................................................................................... Megawatt NEI. .............................................................................................. Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NESP ................. ~ ........................ *......................... National Environmental Studies Project 1 NORAD ...... ********************************************* North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event NPP ............................................*...................................................... Nuclear Power Plant NRC ............. '. .......... :....................................................... Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System OBE ...................................................................................... Operating Basis Earthquake
.......................................................................................................................................... OCJ\ ....................................................... :.............................................. Ovvner Controlled ,A.rea ODCM ............................................................................. Offsite Dose Calculation Manual ONEP ................................................................................... Off-Normal Event Procedure ORO .........................................................................-........ Offsite* Response Organization PA.*.........................................................................-..................................... Protected Area PAG ........................................................................................ Protective Action Guideline PB ...................................................................... :.....................................*........ Pushbutton PCIS ..................................................................... Primary Containment Isolation System Page 22 of 270
I
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX PRA/PSA ..................... Probabilistic Risk Assessment I Probabilistic Safety Assessment PSIG ........................................................... .'.................... Pounds per Square Inch Gauge R .......................................................................................................................... Roentgen RCB ................................................................................................................ RCS Barrier RCIC.: ............................................................................... Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....... 1................................ ................................ Roentgen Equivalent Man RETS .............*............................................ Radiological Effluent Technical Specifications RHR ............................................................................................. Residual Heat Removal RPS ............................ :........................................................... Reactor Protection System RPT .................... J ........**.*****.*.**************.********..******.*.**..*..*.......*... Recirculation Pump Trip RPV ........................................................................................... Reactor Pressure Vessel RWCU .......................................................................................... Reactor Water Cleanup SAP ....................................................................................... Severe Accident Procedure SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ................................ ; ...................................... Self-Contained Breathing Apparatus SOGA .............................................................................. Security Owner Controlled Area SPDS ............................................................................Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SRV ............ ;~ ....................................................................................... Safety Relief Valve SSE ....................................................................................... Sa~e Shutdown Earthquake TEDE ...................................... :........................................ Total Effective Dose Equivalent TAF ........................... ;.. ~ ........................................................................ Top of Active Fuel TSC .......................................................................................... Technical Support Center UFSAR ................................................................... Updated Final Safety Analysis Report USGS ............................................................................ United States Geological Survey Page 23 of 270
~===- Entergy Grand Gulf Nuclear Station EAL Basis,,Document Revision XXX 6.0 GGNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-referen ce is provided to facilitate association and location of a GGNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the GGNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
GGNS NEI 99-01 Rev. 6 EAL Example IC EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 Page 24 of 270
- ~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX GGNS NEI 99-01 Rev. 6 Example EAL IC EAL AG2.1 AG2 1 CUt.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CUS.1 cus 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1
\
CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 1 HU2 1 Page 25 of 270
./
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX GGNS NEI 99-01 Rev. 6 Example EAL IC EAL HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 1 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU? 1 HA1.1 HA1 1, 2 HAS.1 HAS 1 HA6.1 HA6 1 HA7.1 HA? 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS? 1 HG7.1 HG? 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SUS.1 SU4 1, 2, 3 SU6.1 SUS 1 Page 26 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX GGNS N,EI 99-01 Rev. 6 Example EAL IC EAL SU6.2 SUS 2 SU7.1 SU6 1, 2, 3 N/A SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SAS 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SSS 1 SS6.1 SSS 1 SG1.1 SG1 1 SG1.2 SGB 1 Page 27 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Page 28 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category A - Abnormal Rad Levels I Rad Effluent (
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may b~ indicative of a failure of containment systems or precursors to more significant releases. At higher release*rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events ofthis category pertain to the following subcategories:
- 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
- 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
- 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification. ,
Page 29 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer '
EAL: AU1.1 Unusual Event Reading on any Table A-1 effluent radiation monitor> column "UE" for ~ 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE SBGT A/B 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1E+O Ci/sec 6. 7E-2 Ci/sec 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec CTMTVent en 0 5.1E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6.7E-2 Ci/sec Cl) en Radwaste Building Vent
*ns C) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6. 7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent
- E
- , 7.33E+5 cpm er Radwaste --- ---- ----
1
- i Page 30 of 270
I -~Entergy I Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: All I Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by pl_ant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
- B'asis:
This IC addresses a potential reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitm,ents for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features.intended to control the release of radioactive effluents to the envi.ronment. Further, there are administrative controls established to pr~vent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible FICCident events and conditions. - Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such releases ar~ typically associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). Escalation of the emergency classification level would be via IC AA 1. Page 31 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. IAS-04-1-01-017-1 Process Radiation Monitoring
- 2. Offsite Dose Calculation Manual
- 3. XC-Q1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values
- 4. NEI 99-01 AU1 Page 32 of 270
Grand Gulf Nuclear Station EA~ Basis Document Revision XXX Attachment 1 ~mergency Action Level Tec.hnical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - ~adiological Effluent Initiating Condition: Release of gaseo.us or liquid radioactivity greater than 2 times the
. ODCM limits for 60 minutes or longer.
EAL: AUt.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for~ 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): None Basis: This IC addresses a potential reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. 1 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
- Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
Page 33 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).
Escalation of the emergency classification level would be via IC AA 1. Reference(s):
- 1. Offsite Dose Calculation Manual (
- 2. NEI 99-01 AU1 Page 34 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 -, Radiological Efflu~nt
( Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.1 Alert Reading on any Table A-1 effluent radiation monitor> column "ALERT" for ~ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded,1 or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. .Note 2: \ If an ongoing release is detected and the release start time is unknown, assu~e that the release 1 duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, th.en the effluent monitor reading is no longer VALID for classification purposes. Note4 The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessme.nts until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classification Thresholds I Release Point I GE I SAE I Alert I UE I 8.1 E+2 Ci/sec 8.,1 E+1 Ci/sec 8.1 E+O Ci/sec 6.?E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6.?E-2 Ci/sec CTMTVent Ill
- l 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6.?E-2 Ci/sec
- Cl)
Ill Radwaste Building Vent ca I (!J 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6. ?E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.?E-2 Ci/sec Fuel Handling (Aux BLDG) Vent "C
':i 7.33E+5 cpm C' Radwaste --- ---- ----
- J Page 35 of 270
-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
All Definiti~n(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the lev~I of safety of the plant as indicat~d by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). . Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. 1 Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AS1. Reference(s):
- 1. IAS-04-1-01-017-1 Process Radiation Monitoring
- 2. XC-Q1017-1 7001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold V~lues
- 3. NEI 99-01 AA1 Page 36 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision __/ XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of ,.gaseous or liquid radioactivity res~lting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 si,ould be used for emergemcy classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level. of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
- Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AS1. Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AA1 Page 37 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or a.ctual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases .. Releases of this magnitude represent an actual or potential substantial degradation of the level, of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid*CDE. This EAL is assessed per the ODCM (ref. 2) Escalation of the emergency classification level would be via IC AS1. Page 38 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. Offsite Dose Calculation Manual
- 3. NEI 99-01 AA1 Page 39 of 270
--- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory : 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:
AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
- Closed window dose rates > 10 mR/hr expected to continue for~ 60 min.
- Analyses of field survey samples indicate thyroid COE> 50 mrem for 60 min. of inhalation.
(Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability : All Definition(s) : SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent a*n actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Page 40 of 270
--- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AS1.
Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AA1 Page 41 of 270
-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory : 1 - Radiological Effluent.
Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor> column "SAE" for ~ 15 min. (Notes 1, 2, 3, 4) 1 Note 1: The Emergency Director should declare,the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classificatio n Thresholds Release Point GE SAE Alert UE
)
SBGT A/B ,, 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6. 7E-2 Ci/sec CTMT Vent 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec UI
- s 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec G)
Radwaste Building Vent 5.1 E-1 Ci/sec 6.7E-2 Ci/sec UI ca (!) Turbine Building Vent 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6. 7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6. 7E-2 Ci/sec
"'C *; 7.33E+5 cpm C" Radwaste --- ---- ----
- i Page 42 of 270
I Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check,_ or (2) indications orf related or redundant indicators, or (3) by direct observation byplant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that result~ in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs) .. It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established iQ consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established. _If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path;- then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1.
- Reference(s):
- 1. IAS-04-1-01-017-.1 Process Radiation Monitoring
- 2. XC-Q 1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values
- 3. NEI 99-01 AS1
' Page 43 of 270
-:: : :- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid ~DE EAL:
AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available . . Mode Applicability: All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a ba'sis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose)s set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratid of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AG 1. Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AS1 Page 44 of 270 .
"::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:
AS1 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
- Closed window dose rates > 100 mR/hr expected to continue for~ 60 min.
- Analyses of field survey samples indicate thyroid COE > 500 mrem for 60 min. of inhalation.
(Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: AIL Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. 1 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are *associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to prbvide a basis for classifying events and conditions that cannot b~ readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.
- Escalation of the emergency classification level would be via IC AG1.
Page 45 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Techr1ical Bases Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AS1 Page 46 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Bc;:isis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:
AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor> column "GE" for '~ 15 min. (Notes 1, 2, 3, 4) I Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1 E+2 Ci/sec 8.1 E+1 Ci/sec 8.1 E+O Ci/sec 6.7E-2 Ci/sec SBGT AJB I
6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+O Ci/sec 6. 7E-2 Ci/sec CTMTVent Ill 0 5.1 E+1 Ci/sec 5.1 E+O Ci/sec 5.1 E-1 Ci/sec 6. ?E-2 Ci/sec Cl) Ill Radwaste Building Vent ca (!) 1.3E+1 Ci/sec 1.3E+O Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6. 7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent I
~ ':i 7.33E+5 cpm
- C" Radwaste --- ---- ----
- J Page 4 7 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs).
- It includes both monitored and un-'monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have ' stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Reference(s):
- 1. IAS-04-1-01-017-1 Process Radiation Monitoring
- 2. XC-Q1017-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values ,
\ )
- 3. NEI 99-01 AG1 Page 48 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- Mode Applicability:
All Definition(s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiolog.ical effluent EALs more fully addresses the spectrum of possible accident events and_ conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AG1 Page 49 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity re$ulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:
AG1 .3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
- Closed window dose rates > 1,000 mR/hr expected to continue,, for~ 60 min.
- Analyses of field survey samples indicate thyroid COE > 5,000 mrem to*r 60 min. of inhalation.
(Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed.an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release I
)
duration has exceeded the specified time limit. Mode Applicability: All Definitioh{s): SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE. Page 50 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 1O-S-01-12 Radiological Assessment and Protective Action Recommendations
- 2. NEI 99-01 AG1 Page 51 of 270
~
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Levei Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by Fuel Pool Drain Tank low water level alarm, visual observation, water level drop in Upper Ctmt Pools, Aux Bldg Fuel Pools or the Fuel Transfer Canal AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
- Ctmt 209 Airlock (1021 K630)
- Ctmt Fuel Hdlg Area (1021 K626)
- Aux Bldg Fuel Hdlg Area(1021 K622)
Mode Applicability: All Definition(s): UNPLANNED-. A parameter change. or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY- Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway.* Basis: This IC addresses a drop in water level above irradiated fuel sufficient to cause elevated radiation levels.* This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level drop will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel ) (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations. Page 52 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases The effects of plal)ned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or .movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. I Escalation. of the emergency classification level would be via IC AA2. Reference(s):
- 1. 05-1-02-11-8 High Radiation During Fuel Handling
- 2. 04-1-01-021-1 Area Radiation Monitoring System
- 3. UFSAR 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation
- 4. NEI 99-01 AU2
\
Page 53 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent*fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. REFUELING PATHWAY- Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway. Basis: \ This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the REFUELING PATHWAY. These events present radiological safety challenges to plant personnel and are precursors to a release of radioa9tivity to the environment. As such, they represent an actual or potential substantial degra'dation of the level of safety of the *plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU 1.
- This EAL escalates from AU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation.
(e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. *- 1 While an area radiation monitor could detect a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings Page 54 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AS 1. Reference(s):
- 1. NEI 99-01 AA2 Page 55 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND VALID alarm on any of the following radiation monitors:
- Ctmt Vent (P601-19A-G9)
- FH Area Vent (P601-19A-C11)
- Ctmt 209 Airlock (P844-1 A-A 1)
- Ctmt Fuel Hdlg Area (P844-1A-A3)
- Aux Bldg Fuel Hdlg Area (P844-1A-A4)
Mode Applicability: All Definition(s): CONFINEME NT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This EAL addresses events that have caused actual damage to an irradiated fuel assembly. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1. Page 56 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases This EAL addresses a release of radioactive material caused by mechanical damage to
' irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Escalation of the emergency classification level would be via IC AS1. I Reference(s):
. 1. 05-1-02-11-8 High Radiation During Fuel Handling
- 2. 04-1-01-021-1 Area Radiation Monitoring System
- 3. UFSAR 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation
- 4. Offsite Dose Calculation Manual
- 5. NEI 99-01 AA2 Page 57 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
AA2.3 Alert Lowering of spent fuel pool level to 193 ft. (Level 2) on G41 R040A(B) Mode Applicability: All Definition(s): None Basis: This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel ar,d are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC AS1 or AS2. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 92-193 ft. 2 1/a in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 18J~ ft. 2 1/ 8 in. rounded to 183 ft. for readability) (ref. 1). G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148' Control Building in the Lower Cable Spreading Room. Reference(s):
- 1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling
- 2. NEI 99-01 AA2 Page 58 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:
AS2.1 Site Area Emergency Lowering of spent fuel pool level to 183 ft. (Level 3) on G41 R040A(B) Mode Applicability: All Definition(s):
\
IMMINENT - The trajectory of events or conditions is such that an EAL will be n;iet within a relatively short period of time regardless of mitigation or corrective actions. Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it ts included to provide classification diversity. Escalation of the emergency classification level would be via IC AG1 or AG2. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks ' (Level 2 9& 193 ft. 2 1/ 8 in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 18~2 ft. 2 1/ 8 in. rounded to 183 ft. for readability) (ref. 1). G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148' Control Building in the Lower Cable Spreading Room. Reference(s):
- 1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling
- 2. NEI 99-01 AS2 Page 59 of 270
--===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 183 ft. (Level 3) on G41 R040A(B) for ~ 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): None Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 92-193 ft. 2 1/ 8 in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 18~;?; ft. 2 1/s in. rounded to 183 ft. for readability) (ref. 1). G41 R040A(B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148' Control Building in the Lower Cable Spreading Room. Reference(s):
- 1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling
- 2. NEI 99-01 AG2 Page 60 of 270
*~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
AA3.1 Alert Dose r~te > 15 mR/hr in EITHER of the following areas:
- Control Room (SD21-.K600)
- Central Alarm Station (by survey)
Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room is monitored for excessive radiation by one detector, SD21-K600 (ref. 1). The CAS is included in this EAL because of.its importance to permitting access to areas required to assure safe plant operations. While the CAS is in the Control Room Envelope, there are no permanently, installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area. This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substanti.al degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. Page 61 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference{s):
- 1. 06-IC-1021-R-1001 Area Radiation Monitoring Calibration
- 2. NEI 99-01 AA3 Page 62 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels I Rad Effluent Subcategory: ; 3 - Area Radiation Lev~ls Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table A-3 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Control Building 111' SWGR Rms (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A103) 3 Auxiliary Building 93' RHR B Pump Room (1A 105) 3 Auxiliary Building 93' Corridor (1A 101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203) 3
- Auxiliary Building 119' RHR B Pump Room (1A205) 3 Auxiliary ~uilding 119' RCIC Room (1A204) 3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR B Room (1A306, 1A307) 3 Radwaste Building 118' Radwaste Control Room (OR241) 3 Mode Applicability:
3 - Hot Shutdown Page 63 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable. For AA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is* not warranted if any of the following conditions apply:
- The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.
- The higher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g.,
radiography, spent filter or resin transfer, etc.).
- The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.
Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Page 64 of 270
~Entergy Grand Gulf Nucl'ear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
\
The list of plant rooms or areas with entry-related mode app,licability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal. or emergency condition such as emergency* repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). EAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3 (Mode 3 only). Reference(s):
- 1. Attachment 2 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
- 2. NEI 99-01 AA3 Page 65 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category C - Cold Shutdown I Refueling System Malfunction EAL Group:
- Cold Conditions (RCS temperature s 200°F); EALs in this category are applicable only in one or more cold operating modes.
I Category C EALs are directly associated with cold shutdown or refueling system,,safety functions. Given the variability of plant configurations (e.g., systems out-of-:-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay he.at removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refueling, DEF - Defueled). The events of this category pertain to the following subcategories:
- 1. RPV Level RPV water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
- 2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite 1
and offsite power sources for 4.16 KV ESF buses.
- 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions.
- 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
- 5. Loss of Comm unications 1
Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. Page 66 of 270
;"\
~.Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
- 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification.
Page( 67 of 270
~Entergy I Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:
CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Grand Gulf is equipped with multiple RPV water level instruments including: Wide Range, Fuel Zone, Shutdown Range, Upset Range, and Narrow Range (ref. 1). Multiple instruments on different reference and variable legs should be monitored. The Upset Range and Shutdown Range instruments share a common reference leg; therefore, Narrow Range instruments should be routinely monitored. when relying on Shutdown or Upset Range instrument as the primary 'indication. With the plant in .Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 11.4 in. (ref. 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange. Technical Specifications require at least 22 ft 8 in.- of water above the top of the reactor vessel flange in the refueling cavity during refueling operations (ref. 3). The RPV fla~ge is at approximately 212 in. on the Shutdown Range. (ref. 4). This EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent Page 68 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lowe.r RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL recognizes that the minimum required RPV level can change several times during the course of a refueling outage a's different plant configurations and system lineups are* implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified i~ the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3 . . Reference(s):*
- 1. 02-S-01-40 EP Technical Bases
- 2. 05-S-01-EP-2 RPV Control
- 3. Technical Specifications 3.9.6
- 4. 07-S-14-413 RPV Disassembly
- 5. NEI 99-01 CU1 Page 69 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory : 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:
CU1.2. Unusual Event RPV water level cannot be monitored AND EITHER
- UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory
- Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps/Pool
** Drywell equfpment drain sump
- Drywell floor drain ,~ump
- CTMT equipment drain sump
- CTMT floor drain sump
- Suppression Pool
- RHR A, B, C, HPCS, LPCS, RCIC room sumps
- Auxiliary Building floor drain sump Mode Applicability :
4 - Cold Shutdown, 5 - Refueling Definition(s) : UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED-. A parameter changes or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may* be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level (ref. 1). Page 70 of 270 ,
~Entergy. Grand Gulf Nuclear Station .EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref.
2, 3). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses externalito the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be I indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably I rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED everit that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1 ). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. Reference(s):
- 1. 03-1-01 :5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1 H13-P680 Alarm Response Instruction Panel 1H13-P680
- 4. 05-S-01-EP-4 Auxiliary Building Control
- 5. NEI 99-01 CU1 Page 71 of 270
'v ~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: - RPV Level Initiating Condition: . Significant Loss of RPV inventory EAL:
CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < -42 in. (Level 2) Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: The threshold RPV water level of -42 in. is the Level 2 actuation setpoint for HPCS and RCIC. Although RCIC cannot restore RPV inventory in the cold condition, the Level 2 actuation setpoint is operationally significant and is indicative of a loss of RPV inventory significantly below the low RPV water level scram setpoint specified in CU1 .1 (ref. 1, 2). This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery. Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):
- 1. Technical Specifications Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation
- 2. 04-1-02-1H13-P601-16A-A4Alarm Response Instruction Panel 1H13-P601 panel 16A-A4 RX LVL 2 (-42") LO
- 3. NEI 99-01 CA 1 Page 72 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases . Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for ~ 15 min. (Note 1) AND EITHER
- UNPLANNED rise in any Table C-1 sump or pool Jevel due to a loss of RPV inventory
- Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, orwill likely be exceeded. The Emergency Director is n'ot allowed an additional 15 minutes to declare after the time limit is exceeded.
Table C-1 Sumps/Pool
- Drywell equipment drain sump
- Drywell floor drain sump
- CTMT equipment drain sump
- CTMT floor drain sump
- Suppression Pool
- RHR A, B, C, HPCS, LPCS, RCIC room sumps
- Auxiliary Building floor drain sump Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): UNISOLABLE -An open or breached system line that cannpt be isolated, remotely or locally. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause ?f the p~rameter change or event may be. known or unknown. Page 73 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachmen t 1 Emergenc y Action Level Technical Bases Basis:
In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2, 3). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containme nt from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplaine d rise in suppressio n pool' water level could be . indicative of RHR valve misalignme nt or leakage. If the make-up rate to the RPV unexplaina bly rises above the pre-establi shed rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediate ly identified. Visual observatio n of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, the inability to monitor RPV level may be caused by instrument ation and/or power failures, or water level dropping below the range of available instrument ation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1. If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. - Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H 13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1 H13-P680 Alarm Response Instruction Panel 1H13-P68 0
- 4. 05-S-01-E P-4 Auxiliary Building Control
- 5. NEI 99-01 CA 1 Page 74 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss .of RPV inventory affecti.ng core decay heat removal capability EAL:
CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV water level< -150 in. (Level 1) Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s ): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The threshold RPV water level of -150 in. is the low-low-low ECCS actuation setpoint (Level 1). The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further lowering of RPV water level and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier (ref. 1, 2). This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Followin,g an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. Outage/shutdown ,contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control Page 75 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases functions. The difference in the specified RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal,* SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s):
- 1. Technical Specifications Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation
- 2. 04-1-02-1H13-P601-17A-E2 Alarm RespoF1se Instruction Panel 1H13-P601 panel 17A-E2 RX LVL 1 (-150") LO
- 3. NEI 99-01 CS1 Page 76 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcate gory: 1 - RPV Level 1 \ Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV water level< -167 in. (TAF) I Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of-167 in.), core uncovery starts to occur (ref. 1). This IC addresses a significant and prqlonged loss of RPV level control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probabl.e. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission
- product release to the environment.
Page 77 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the emergency classification level would be via IC CG1 or AG1. Reference(s):
- 1. 02-S-01-40 EP Technical Bases
- 2. NEI 99-01 CS1 Page 78 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condijtion: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .3 Site Area Emergency. RPV level cannot be monitored for~ 30 min. (Note 1) AN[} Core uncovery is indicated by any of the following: I
- UNPLANNED rise in Suppression Pool l~vel of sufficient magnitude to indicate core uncovery
- Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
- Containment/Drywell High Range Area Radi~tion Monitor (1 D21-K6481\K648B-GC)
~ 100 R/hrhigh high alarm Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
- Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
\
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. -) UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to *indicate water level in the refuel cavity and the Core Plate d/p ipstrument which *is re-spanned and re-scaled to indicate water level (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must 'be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are Page 79 of 270
, -===- Entergy
- Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in Suppression Pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV une.xplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory.
In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise, with corresponding indications on area radiation monitors. 100 R/hr is used for this indication A high high alarm(> 1,000 R/hr) on Containment High Range Radiation Monitors (1 D21-K648B and Cl, or associated computer or recorder points (Ref 6.)) is selected as the control room alarm indication to provide this function. These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity elevation.
- In Cold Shutdovm Mode, 'Nhen the core is uncovered, the dose rate in the containment or dry'Nell around or above the core 'Nill rise, with corresponding indications on area *radiation monitors. A high high alarm (> 1,000 R/hr) on Ory\vell/Containment High Range Radiation Monitors (1021 K648A, B, C, 0, or associat~d computer or recorder points (Ref 6.)) is selected as the control room alarm indication to provide this function. The detectors are located in the drywell and the containment and are in a position to monitor the dryvlell and containment direct radiation environment.
This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Fqllowing an_,extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. In this EAL, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monit.or RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential source~ of wate~ flow to ensure they are indicative of leakage from the RPV. Page 80 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
'Attachment 1 Emergency Action Level Technical Bases I
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power.Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NU MARC \ 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or AG1 Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1--02-1 H13-P680 Alarm Response Instruction Panel 1H13-P680
- 4. 05-S-01-EP-4 Auxiliary Building Control
- 5. 06-IC-1021-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration
- 6. NEI 99-01 CS1 7., Calculation J-021-1, Set Points Determination For High Range DW & Containment Radiation Monitors (021 System)
Page 81 of 270
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~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory : 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged
- EAL:
CG1 .1 General Emergency RPV level< -167 in. (TAF) for;::: 30 min. (Note 1) AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General. Emergency is not required. Table C-2 Containmen t Challenge Indications
- CONTAINMENT CLOSURE not established (Note 6)
- Drywell or containment hydrogen concentration > 4%
- UNPLANNED rise in containment pressure
- Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)
Mode Applicability : 4 - Cold Shutdown, 5 - Refueling Definition(s) : CONTAINME NT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Page 82 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may .be known or unknown.
Basis: When RPV level drops below -167 in., core uncovery starts to occur (ref. 1). Four conditions are associated with a challenge to Containment integrity:
- CONTAINMENT CLOSURE is not established.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a
- core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen bum (4%). The Igniter S¥stem is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as* the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition.
Following ignition, hydrogen is consumed through formation of di.ffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 2).
- Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
- Secondary Containme.nt radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 3).
This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss ofcontainment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-II Page 83 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
\ 'I Attachment 1 Emergency Action Level Technical Bases established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixtl:Jre in containment. If all installed hydrogen gas monitors are out-of-service- during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentra1tion reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Reference(s):
- 1. 02-S-01-40 EP Technical Bases
- 2. BWROG Emergency Procedure and Severe Accident Guidelines, Revision 3, p. B-16-64
- 3. 05-S-01-EP-4, Auxiliary Building Control
- 4. NEI 99-01 CG1 Page 84 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases j
Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL: CG1 .2 General Emergency RPV level cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any ofthe following:
- UNPLANNED rise in Suppression Pool level of sufficient magnitude to indicate core uncovery
- Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core ur:icovery
- Containment/Drywell High Range Area Radiation Monitor (1 D21-K648AK648B-GC)
~ 1OOR/hr high high alarm AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, d~claration of a General Emergency. is not required. *
- Table C-2 Containment Challenge Indications
- CONTAINMENT CLOSURE not established (Note 6)
- Drywell or containment hydrogen concentration > 4%
- UNPLANNED rise in containment pressure
- Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)
Page 85 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: - In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level (ref. 1). In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplai'ned rise in Suppression Pool water level could be indicative of RHR valve misalignment o,r leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise, with corresponding indications on area radiation monitors. 100 R/hr is used for this indication A high high alarm(> 1,000 R/hr) on Containment High Range Radiation Monitors (1 D21-K648B and Cl, or associated computer or recorder points (Ref 6.)) is selected as the control room alarm indication to provide this function. These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity elevation. *
- Page 86 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Lev,el Technical Bases Four conditions are associated with a challenge to Containment integrity:
- CONTAINMENT CLOSURE is not established.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen bum (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition.
Following ignition, hydrogen is c.onsumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 4).
- Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capabili.ty. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission* product release.
- Secondary Containm~nt radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 5).
This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high*potential for a direct and unmonitored reiease of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is suffici~nt to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral eqdipment Page 87 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachmen t 1 Emergenc y Action Level Technical Bases damage leading to a loss of containment integrity. It therefore represents a challenge to Containme nt integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containme nt will preclude personnel access. During periods when installed containme nt hydrogen gas monitors are* out-of-service, operators may use the other listed indications to assess whether or not containme nt is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performan ce of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Powe r Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Asse~s Shutdown Management. Reference(s):
- 1. 03-1-01-5 Refueling
- 2. 04-1-02-1 H13-P601 Alarm Response Instruction Panel 1H13-P601
- 3. 04-1-02-1 H 13-P680 Alarm Response Instruction Panel 1 H 13-P680
- 4. BWROG Emergenc y Procedure and Severe Accident Guidelines, Revision 3, p. B-16-64
- 5. 05-S-01-EP-4, Auxiliary Building Control
- 6. 06-IC-102 1-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration
- 7. NEI 99-01 CG1
- 8. Calculation J-021-1, Set Points Determination For High Range OW & 'Containment Radiation Monitors (021 System)
Page 88 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 - Loss of ESF AC Power Initiating Condition: Loss of all but one AC power source to ESF buses for 15 minutes or longer EAL:
CU2.1 Unusual Event AC power capability, Table C-3, to DIV I and DIV II ESF 4.16 KV buses reduced to a single po,wer source for~ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-3 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
(
- DIV I DG (DG 11)
- DIV II DG (DG 12)
Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant-operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 89 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout.
- This IC describes a significant degradation of offsite and onsite AC power sources such that any ~dditional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an_ Alert because of the greater time:available to restore another power source to service. Additional time is available-,due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.
- A loss of all offsite power with a concurrent failure of all but one emergency ESF power source (e.g., an onsite diesel generator).
- A loss of all offsite power and loss of all emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being back-fed from the unit main generator.
- A loss of emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. This EAL is the cold condition equivalent of the hot condition EAL SA 1.1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 CU2 Page 90 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C ~ Cold Shutdown I Refueling System Malfunction Subcategory: 1 2 - Loss of ESF AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer EAL:
CA2.1 Alert Loss of all offsite and all onsite, AC power to DIV I and DIV II ESF 4.16 KV buses for~ 15 min. (Note 1) I Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 - C9ld Shutdown, 5 - Refueling, bEF - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the*reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basi,s accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. Page 91 of 270
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore an ESF bus to service.
Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. 1Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of the hoticondition EAL SS1 .1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 CU2 Pag,e 92 of 270
~.Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases I
Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3.- RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability: I 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and ,components as a functional barrier to fission
- product release under shutdown conditions.
Containment Closure is escablished when either Primary or Secondary Containment integrity is established. I UNPLANNED~. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F) (ref. 1, 2). In the absence of reliable RCS temperature .indication, classification is based on the concurrent loss of reactor vessel level indications *per EAL CU3.2. This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the
\,
level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1. A momentary UNPLAN.NED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS .in excess ofthat which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Page 93 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel I flange are carefully planned and Gontrolled. A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.
\
Escalation to Alert would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s):
- 1. Technical Specifications Table 1.1-1
- 2. 03-1-01-3 Plant Shutdown
- 3. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal 4 ..NEI 99-01 CU3 Page 94 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL:
CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has i been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 - Cold Shutdown, 5- Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barri~r to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This EAL addresses the inability to determine RCS temperature ano level; and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSUR~ is not established during this event, the Emergency Director 1 should also refer to EAL CA3.1. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. 1 Escalation to Alert would b~ via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
- Page 95 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Reference(s):
- 1. 02-S-01-40 EP Technical Bases
- 2. Technical Specifications Table 1.1-1
- 3. 03-1-01-3 Plant Shutdown
- 4. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal
- 5. NEI 99-01 CU3 Page 96 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED rise in RCS tempera1ture to > 200°F for> Table C-4 duration (Note 1) OR UNPLANNED RPV pressure rise > 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-4 RCS .Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status Intact N/A 60 min.* established 20 min.* Not intact not established O min.
- If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Mode Applicability: 4 - Cold Shutdown, 5 *- Refueling Definition(s): CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when either Primary or Secondary Containment integrity is established. UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 97 of 270
,::::::- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
. Attachment 1 Emergency Action Level Technical Bases Basis:
In the absence of reliable RCS temperature indication, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4. This EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise. , The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with th,e RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature ri~e without a substantial degradation in plant safety. Finally, in the case where there is a rise in RCS temperature, the 1RCS is not intact and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The RCS pressure rise threshold provides a pressure-based indication of RCS.heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or AS1. Reference( s): 1 Technical Specifications Table 1.1-1
- 2. 03-1-01-3 Plant Shutdown
- 3. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal
- 4. NEI 99-01 CA3 Page 98 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX 1.
Attachment 1 .Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:* CU4.1 Unusual Event Indicated voltage is < 105 VDC on required vital 125 VDC buses 11.DA and 11 DB for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has
- been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis Vital DC buses 11 DA and 11 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1. 72 volts/cell. These. cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC) (ref. 1, 2). This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions raise the .time available to restore a vital DC bus 1to service. Thus, this condition is considered to be a p9tential degradation of the level of safety of the plant. As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Page 99 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
\
Attachment 1 Emergency Action Level Technical Bases Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA 1 or CA3, or an IC in Recognition Category A. This EAL is the cold condition equivalent of the hot condition EAL SS2.1. Reference(s):
- 1. Calculation No: EC-01111-14 001 Station Division I Battery 1A3 and Division II Battery 183 Discharge Capacity during Extended Loss of AC Power
- 2. UFSAR 8.3.2.1.1 Station DC Power
- 3. NEI 99-01 CU4 Page 100 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revis'ion XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods State/ System Onsite NRC Local Station Radio System x GGNS Plant Phone System x Public Address System x Emergency Notification System (ENS) x Commercial Telephone System x x Satellite Phones x x INFG~M NetifieatieR System :X Operational Hotline x Mode Applicabilify: 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s): None Page 101 of 270
-=-=~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the N RC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriff's Department, Louisiana Department of Environmental Quality, Tensas Parish Sheriff's Office, and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness. The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the cold condition equivalent of the hot condition EAL SU7.1. Reference(s):
- 1. GGNS Emergency Plan Section 7.5, Communications Systems
- 2. 04-S-01-R61-1 Plant Communications
- 3. NEI 99-01 CU5 Page 102 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety ~ystems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode '
EAL: CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
- Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
- Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table C-6 Hazardous Events
- Seismic event (earthquake)
- Internal or external FLOODING event
- High winds or tornado strike
- FIRE
- EXPLOSION
- Other events with similar hazard characteristics as determined by the Shift Manager Page 103 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Mode Applicability:
4 - Cold Shutdown, 5 - Refueling Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): I Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: * (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. *, VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; Page 104 of 270
\ -~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should* be significant enough to cause concern regarding the operability or reliability of the SAFETY .SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of the hot condition EAL SA8.1. Reference(s):
- 1. EP FAQ 2016-002
- 2. NEI 99-01 CA6 Page 105 of 270
-~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. .
- An Unusual Event is declared on the basis of the occurrence of an event of suffici~nt magnitude that a loaded cask CONFINEMENT BOUNDAR'Y is damaged or violated.
The GGNS ISFSI is located wholly within the plant PROTECTED AREA Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs. ( Page 106 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack)
> EITHER of the following:
- 60 mrem/hr (gamma + neutron) on the top of the overpack
- 600 mrem/hr (gamma+ neutron) on the side of the overpack (excluding inlet and outlet ducts)
Mode Applicability: All Definition{s): CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). INDEPENDENT SPENT FUEL STORAGE INSTALLATION (/SFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Basis: This IC addresses an event that results in c;tamage to the CONFINEMENT BOUNDARY of a . storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values. The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions (ref. 2). The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose Page 107 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases rate. It is recognized that in the case of extreme damage to a loaded_ cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSls are covered under ICs HU1 and HA1. Reference(s):
- 1. UFSAR 9.1.4.2.10.4 Storage of Fuel at the Independent Spent Fuel Storage Installation
- 2. GGNS HI-STORM 100 10 CFR 72.212 Evaluation Report Licensing Basis Document, Revision 10, Section 4.2.4 (Section 5.7) Radiation Protection Program
- 3. NEI 99-01 E-HU1 Page 108 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category F - Fission Product Barri.er Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves. G. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds
*are used as criteria for escalation of the. Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any Joss or any potential Joss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential Joss of any two barriers General Emergency: Loss of any two barriers and Joss or potential Joss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
- The Fuel Clad Barrier and the RCS Barher are weighted more heavily than the Containment
( Barrier. Page 109 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases
- Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
- For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded.
- The fission product barrier thresholds specified within a scheme reflect plant-specific GGNS design and operating characteristics.
- As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.
The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.
- At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
Page 110 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:
FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS bar,rier (Table F-1) Mode Applicability: 1 -Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1 Reference(s):
- 1. NEI 99-01 FA 1 Page 111 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: NIA Initiating Condition: Loss or potential loss of any two barriers EAL:
FS1 .1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
- One barrier loss and a second barrier loss (i.e., loss - loss)
- One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
- One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss) At the 1Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT. Reference(s):
- 1. NEI 99-01 FS1 Page 112 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency A~tion Level Technical Bases I
Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table _F-1 lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
- Loss of Fuel Clad, RCS and Containment Barriers
- Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
- Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
- Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference(s):
- 1. NEI 99-01 FG1 Page 113 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss* thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are: A. RPV Water Level B. RCS Leak Rate C. Containment Conditions D. Containment Radiation I RCS Activity E. Containment Integrity or Bypass ( F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2 ... FCB6). If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed .all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss threshold.s. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Page 114 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 Emergency Action* Level Technical Bases Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and FA 1.1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then 8, ... , F. Page 115 of 270
*-:::::: Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB)
Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A FCB1 SAP entry is required FCB2 RPVwater level cannot be restored and maintained RCB1 RPVwater level cannot be restored and maintained RPVWater > -167 in. (TAF) None None CNB1 SAP entry is required
> -167 in. (TAF)
Level or cannot be determined or cannot be determined
- RCB2 UNISOLABLE break in any of the following:
. Main steam line RCB4 UNISOLABLE primary system leakage that results in exceeding CNB2 UNISOLABLE primary system leakage that results in EITHER: exceeding EITHER:
B RCIC steam Line None None One or more EP-4 radiation One or more EP-4 MAX RWCU RCS Leak Rate Feedwater HPCS
. Operating Limits One or more EP-4 area temperature Operating
. SAFE area radiation levels One or more EP-4 MAX SAFE area temperature None RCB3 Emergency Depressurization Limits levels is required CNB5 Containment pressure > 15 psig CNB3 UNPLANNED rapid drop in c None None RCB5 Drywell pressure> 1.23 psig None containment pressure following containment pressure rise CNB6 Drywell or containment hydrogen concentration > 4%
CTMT due to RCS leakage CNB7 Parameters cannot be restored CNB4 Containment pressure Conditions response not consistent with and maintained within the safe LOCA conditions zone of the HCTL curve (EP Figure 1) FCB3 Containment radiation (RITS-D K648B or C) > 400 R/hr RCB6 Drywell radiation (RITS-K648A CNB8 Containment radiation (RITS-CTMT Rad/ FCB4 Primary coolant activity None None None or D) > 100 R/hr K648B or C) > 7000 R/hr RCS > 300 µCi/gm dose Activity equivalent 1-131
- CNB9 UNISOLABLE direct E downstream pathway to the environment exists after CTMT None None None None Containment isolation signal None Integrity or Bypass CNB101ntentional Containment venting per EPs F FCB5 Any condition in the opinion FCB6 Any condition in the opinion of RCB7 Any condition in the opinion of RCB8 Any condition in the opinion of CNB11 Any condition in the opinion of CNB12Any condition in the opinion of of the Emergency Director the Emergency Director that the Emergency Director that the Emergency Director that the Emergency Director that the Emergency Director that Emergency that indicates loss of the fuel indicates potential loss of the indicates loss of the RCS indicates potential loss of the indicates loss of the indicates potential loss of the Director clad barrier fuel clad barrier barrier RCS barrier Containment barrier Containment barrier Judgment Page 116 of 270
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~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1. - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:
I FCB1 SAP entry is required Definition(s): None Basis: Emergency Procedures (EPs) specify entry to the Severe Accident Procedures (SAPs) when c0re cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining,RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2).
- The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.
1 This threshold is also a Potential Loss of the Containment barrier (CNB1 ). Since SAP entry qccurs after core uncovery has occurred a Loss .of the RCS barrier exists (RCB1 ). SAP entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. The Loss threshold represents the EOP requirement for entry into the SAPs. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Reference(s):
- 1. 05-S~01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. EP FAQ 2015-004'
- 4. NEI 99-01, RPV Water Level Fuel Clad Loss 2.A Page 117 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:
FCB2 RPV water level cannot be restored and maintained> -167 in. (TAF) or cannot be determined Definition(s): _ None Basis: An RPV water level instrument reading of-167 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e.1, steam cooling or spray cooling). If core uncovery is threatened, the EPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier. When RPV water level cannot be determined, EPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the fuel clad barrier is th'reatened and reliance on alternate means -of assuring adequate core cooling must be attempted. The instructions in EP-5 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification. Page 118 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases This water level corresponds to the top of the active fuel and' is used in the EOPs to indicate a challenge to core cooling.
The RPV water level threshold is the same as RCS barrier Loss threshold RCB1. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the spe'cified level following depressurization of the RPV (either manually, aut9matically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level l control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. I Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified. Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. 05-S-01-EP-2A ATWS RPV Control 4 NEI 99-01 RPV Water Level Potential Loss 2.A Page 119 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:
I None Page 120 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:
I None Page 121 of 270
'"Entergy ~ -===-- Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Loss Threshold:
I None Page 122 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold:
I None Page 123 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Loss Threshold:
FCB3 Containment radiation (RITS-K648B or C) > 400 R/hr Definition(s): None Basis: The containment radiation monitor reading (425 R/hr rounded to 400 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 1.6% fuel clad damage (ref. 1). Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold RCB6 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency. There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation I RCS Activity. Reference( s):
- 1. XC-Q1021-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values
- 2. 04-1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A
\_
Page 124 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Loss Threshold:
FCB4 Coolant activity> 300 µCi/gm.dose equivalent 1-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hour~ to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is* no Fuel Clad barrier Potentiql LOss threshold associated with CTMT Radiation I RCS Activity. Reference(s):
- 1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A Page 125 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation I RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 126 of 270
r
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment,1 - Emergency Action Level Technical Bases 1
\
Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: I None Page 127 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass 1 Degradation Threat: Potential Loss Threshold: I None Page 128 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX '
Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: FCB5 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad Barrier Definition(s): None rBasis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. Reference(s):. 1.. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 129 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:
FCBG Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad Barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A ,
Page 130 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RPV Water Level qegradation Threat: Loss
\
Threshold:
- RCB1 RPV water level cannot be restored and maintained> -167 in. (TAF) or cannot be determined DJfinition(s):
None. Basis: An RPV water level instrument reading of-167 in. indicates level is at the top of.active fuel (TAF) (ref. 1). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Containment barriers, and initiation of all ECCS. If RPV water level cannotbe maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the lowering level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier. When RPV water level cannot be determined, EOPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). The instructions in EP-5 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss RCB3). Note that EP-2A, A TWS RPV Control, may require intentionally lowering RPV water level to -167 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - RPS Failure EALs.
- In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top ofactive fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification.
This water level corresponds to the top of active fuel and is used in the EOPs to indicate a challenge to core coolin.g. ' Page 131 of 270
-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold FCB2. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area 1
Emergency. , This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV
- water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergeney RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) rJO low*
pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. There is no RCS barrier Potential Loss threshold associated with RPV Water Level. Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. 05-S-01-EP-2A ATWS RPV Control
- 4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 132 of 270
/ -====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Cool~nt System Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold:
I None Page 133 of 270
-~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision xixx
'Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak.Rate Degradation Threat: , Loss Threshold:
RCB2 UNISOLABLE break in any of the following:
- Main steam line
- RCIC steam line
- RWCU
_*_Feedwater
- HPCS Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside containment (e.g., EP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see Loss CNB9) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do not contain steam, they are included in the list . because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. Page 134 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Even though the High Pressure Core Spray (HPCS) injects into the RCS, it is included in this EAL due to the potential for an inter-system loss of coolant back flowing from the discharge lines (via failed isolation valves and check valves) and out through a break in the piping. A HPCS failure that does not result in back flow of RCS and out through a break should not be considered as meeting the EAL threshold.
Large high-energy lines that rupture outside containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, remotely or locally,, the RCS, barrier Loss threshold.is met. Reference(s):
- 1. NEI 99-01 RCS Leak Rate RCS Loss 3.A Page 135 of 270
-===- Entergy Grand Gul*f Nuclear Station EAL Basis Docume~t Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:
RCB3 Emergency Depressurization is required Definition(s): None Basis: Emergency Depressurization in accordance with the EOPs (ref. 1, 2) is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. EP-2 RPV Control - Emergency Depressurization allows terminating (he depressurization if necessary to maintain RCIC as an injection source. This would require closing the SRVs. Even though the SRVs may be reclosed, this threshold is still met due to the requirement for an Emergency Depressurization having been met (ref. 2). Reference(s):
- 1. 05-S-01-EP-2 RPV Control - Emergency Depressurization
- 2. EP FAQ 2015-003
- 3. NEI 99-01 RCS Leak Rate RCS Loss 3.8 Page 136 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System I Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:
- One or more EP-4 area radiation Operating Limits
- One or more EP-4 area temperature Operating Limits Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of UNISOLABLE primary system leakage outside the containment. The EP-4 entry condition values define this RCS threshold because they are the Operating Limits (maximum normal operating values) and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or mis-operation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1). In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Auxiliary Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e:g. room FLOODING, high area temperatures, reports of steam in the secondary contc;:1inment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment. Potential loss of RCS based on primary system leakage_outside the containment is determined from EOP temperature or radiation EP-4 Operating Limits (Max Normal Operating values) in Page 137 of 270
-=-==- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX AttachmenH - Emergency Action Level Technical Bases areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside containment.
An EP-4 Operating Limit value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a reduction in the steam or water being discharged through an unisolated break in the system.
,, I An UNISGLABLE leak which is indicated by EP-4 Operating Limit values escalates to a Site Area Emergency when combined with Containment Barrier Loss thresholds CNB 2 or CNB9 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
Reference(s):
- 1. 05-S-01-EP-4 Auxiliary Building Control
- 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 138 of 270
~Entergy Grand Gulf Nuclear Station EAL, Basis Document Revision XXX Attachm~nt 1 - Emergency Action Level Technical Bases Barrier: Re,actor Coolant System Category: C. CTMT Conditions Degradation Threat: Loss Threshold:
RCB5 I Drywell pressure > 1.23 psig due to RCS leakage Definition(s): None Basis: I The drywell high pressure scram setpoint is an entry condition to EP-2, RPV Control, and EP-3, Containment Control (ref. 1, 2). Normal containment pressure control functions (e.g.,
- operation of drywell and containment cooling, vent using containment vessel purge, etc.) are specified in EP-3 in advance of less desirable but more effective functions (e.g., operation of containment sprays, etc.).
- In the design basis, containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release, into the containment for which normal pressure control systems are inadequate or incapable of reversing the rising pressure trend.
Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control containment vent/purge (ref. 3). The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect containment pressure. Drywell pressure greater than 1.23 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventocy) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.23 psig should not be considered an RCS barrier Loss. The 1.23 psig value is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system. There is no RCS barrier Potential Loss threshold associated with CTMT Conditions. Page 139 of 270
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~
*===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
\
Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-2 RPV Control
- 2. 05-S-01-EP-3 Containment Control
- 3. UFSAR Section 6.2.1, Containment Functional Design
- 4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A Page 140 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases ,Barrier: Reactor Coolant System Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold:
I None
\
Page 141 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System r Category: D. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:
RCB6 Drywell radiation (RITS-K648A or D) > 100 R/hr Definition(s): None 1 Basis: The drywell radiation monitor reading (150 R/hr rounded to 100 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits (ref. 1). This value is lower than that specified for Fuel Clad Barrier Loss threshold FCB3 since it indicates a loss of the RCS Barrier only. There is no RCS barrier Potential Loss threshold associated with CTMT Radiation/ RCS Activity. Reference(s):
- 1. XC-Q1021-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values . (
- 2. 04-1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 142 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 143 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E. CTMT Integrity or Bypass Deg,radation Threat: Loss Threshold:
I None Page 144 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: I None I j Page 145 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:
RCB7 Any condition in the opinion of the Emergency Director that indicates loss of the RCS Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 146 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Reactor Coolant S1ystem Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:
RCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS Barrier Definition(s ): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):
- 1. N,EI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 147 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Loss Threshold:
I None Page 148 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold: I CNB1 SAP entry is required
/
Definition(s): None Basis: EPs specify entry to the SAPs when core cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2). The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling withoL:,Jt further degradation of the clad. This threshold is also a Loss of the Fuel Clad barrier (Loss FCB1 ). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (Loss RCB1 ). SAP entry,. therefore, represents a Loss of two barriers and a Potential Loss' of a third; which requires a General Emergency classification. The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold FCB1. The Potential Loss requirement for entry into the SAGs indicates adequate core cooling cannot be assured and that core damage/ is possible. BWR EPGs/SAGs specify the conditions when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and greater potential for containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. There is no Containment barrier Loss threshold associated with RPV Water Level. Page 149 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP~2 RPV Control
- 2. 05-S-01-EP-5 RPV Flooding
- 3. EP FAQ 2015-004
- 4. NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 150 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX I
Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:
- One or more EP-4 MAX SAFE area radiation levels
- One or more EP-4 MAX SAFE area temperature levels Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.,* 'UNISOLABLE -An open or breached*~ystem line that cannot be isolated, remotely or locally. I Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isojated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISdLABLE primary system leakage outside the containment. The MAX SAFE values define this Containment barrier threshold because they are indicative of problems in the Secondary Containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1). In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area temperature condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by a fire or loss of area cooling. Conversely, a high area temperature condition in conjunction with other indications (e.g. room FLOODING, high area radiation. levels, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. The Max Safe area temperature values and the Max Safe area radiation values are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the Page 151 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
There is no Containment barrier Potential Loss threshold associated with RCS Leak Rate. Reference( s):
- 1. 05-S-01-EP~4 Auxiliary Building Control
- 1. NEI 99-01 RCS Leak Rate PC Loss 3.C Page 152 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX I
Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: , Potential Loss Threshold: I None Page 153 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: / C. CTMT Conditions Degradation Threat: Loss Threshold:
CNB3 UNPLANNED rapid drop in containment pressure following containment pressure rise Definition(s): UNPLANNED - A parameter change or an event that is not 1) the* result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. I Basis: Rapid UNPLANNED loss of containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity. This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass con'dition. Reference(s):
- 1. NEI 99-01 Primary Containment Conditions PC Loss 1.A Page 154 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technicai Bases Barrier: Containment Catego,ry: C. CTMT Conditions Degradation Threat: Loss I Threshold: CNB4 Containment pressure response not consistent with LOCA conditions Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Containment pressure should rise as a result of mass and energy release into the containment from a LOCA. Thus, containment pressure not rising under these conditions indicates a loss of containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Reference(s):
- 1. USAR Table 6.2-5, Summary of Short Term Containment Responses to Recirculation Line and Main Steam Line Breaks
- 2. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design
- 3. NEI 99-01 Primary Containment Conditions PC Loss 1.8 Page 155 of 270
(
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 ~ Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss.
Threshold: CNB5 Containment pressure > 15 psig Definition(s): None Basis: When the containment pressure exceeds the maximum allowable value (15 psigY (ref. 1), containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). This pressure is based on the containment design pressure as identified in the accident analysis. If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred. The threshold pressure is the containment internal design pressure. Structural acceptance testing demonstrates the capability of the containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Reference(s):
- 1. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design
- 2. 05-S-01-EP-3 Containment Control
- 3. NEI 99-01, Primary Containment Conditions PC Potential Loss 1.A Page 156 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold:
CNB6 Drywell or containment hydrogen concentration > 4% Definition(s ): None Basis: In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, . containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 1). If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the containment, loss of the Containment barrier could occur. Reference(s):
- 1. 02-S-01-40 EP Te.chnical Bases (EP-3 step H-3)
- 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.8 Page 157 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold:
CNB7 Parameters cannot be restored and maintained within the safe zone of the HCTL curve (EP Figure 1) Definition(s): None Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
- Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
- Suppression chamber pressure above Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for' the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. The term "cannot be restored and maintained withinabove " means the parameter value(s) is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to the parameter value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained within a specified limit does not require immediate action simply because the current value is outside the limit, but does not permit extended operation outside the limit; the threshold must be considered reached as soon as it is apparent that operation within the limit cannot be attained. Page 158 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C Page 159 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:
[ None Page 160 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency .Action Level Technical Bases Barrier: Containment Category: 0. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:
CNB8 Containment radiation (RITS-K648B or C) > 7,000I R/hr Definition(s): None Basis: In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (Loss RCB6) and a loss of the Fuel Clad barrier (Loss FCB3) have already occurred. This threshold, therefore, represents a General Emergency classification. The containment radiation monitor reading (7,350 R/hr rounded to 7,000 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming* that 20% of the fuel cladding has failed (ref. 1). This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and R.CS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nucl~ar Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this I condition to exist, there must already have been a loss of the RCS Barrier and the Fu,el Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. There is no Containment barrier Loss threshold associated with CTMT Radiation/RCS Activity. Reference(s):
- 1. XC-Q1021-17001.Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values
- 2. 04-1-01-021-1 SOI Area Radiation Monitoring System
- 3. NEI 99-01 NEI 99-01 Primary Containment Radiation Potential Loss 4.A Page 161 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
CNB9 UNISOLABLE direct downstream pathway to the environment exists after Containment isolation signal Definition(s): UN/SOLABLE - An open_ or breached system line that cannot be isolated, remotely or locall~. VALID -An indication, report, or condition, is considered to be valid when it is v~rified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removep. Implicit in this definition is the need for timely assessment. Basis: Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate crassification. This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity. This threshold also applies to a containment bypass due to a HPCS or LPCS line break outside containment with injection check valve failure allowing an UNISOLABLE direct pathway for RCS release to the environment. The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems' or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE main steam line or RCIC steam line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the main condenser is available with an UNISOLABLE main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the Page 162 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases environment. These minor releases are assessed using the Category A, Abnormal Rad Release I Rad Effluent, EALs.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter coulg become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retent,dn ability has been exceeded) or water saturation from steam/high humidity in the release stream. EP-3 Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a VALID containment isolation signal, the Containment barrier should .be considered lost. Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the ~ecognition Category A ICs. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Reference(s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 163 of 270
1-:->
~ -===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
CNB10 Intentional Containment venting per EPs Definition(s): None Basis: EP-3, Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded. The threshold is met when the operator begins venting the containment in accordance with 3, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 1). Intentional venting of containment for containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Reference( s):
- 1. 05-S-01-EP-3 Containment Control
- 2. NEI 99-01 CTMT Integrity or Bypass Containment Loss 3.B Page 164 of 270
**~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
I None Page 165 of 270
*===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: F. Emergency Director JudgmentI Degradation Threat: Loss Threshold:
CNB11 Any condition in the opinion of the Emergency Director that indicates loss of the Containment Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 166 of 270
-:::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Barrier: Containment Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:
CNB12 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment Barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the.Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status 9annot be monitored. Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 167 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases
. Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
- 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
- 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
- 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
- 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown
- 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.
- 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
- 7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.
Page 168 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards Subcategory: . 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:
HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by GGNS Security Shift Supervision OR Notification.of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands wl be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OGA) For the purposes of classification, the Security area behveen the OCA detection fence and the PROTECTED /\REA boundary knovm as the Security Owner Controlled Area (SOGA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Page 169 of 270
~ ~~ ~ *===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threaUcompromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA 1 and HS1. Timely and accurate communications between Security Shift Supervision and the Control Room is essenti 91 for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Security Plan for GGNS. The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Page 170 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NCDRAD through the NRC. Validation of the threat is performed in accordance with 11-S-82-1 Security Contingency Events (ref. 2).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to. a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1). Escalation of the emergency classification level would be via IC HA1. Reference(s):
- 1. GGNS Security Plan
- 2. 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HU1 Page 171 of 270
{.
-=::~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachmen t 1 - Emergenc y Action Level Technical Bases Category: H - Hazards Subcateg ory: 1 - Security Initiating Condition : HOSTILE ACTION within the SECURITY OWNER CONTROL LED AREA or airborne attack threat within 30 minutes EAL:
HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROL LED AREA as reported by GGNS Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicabil ity: All Definition( s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGE S, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECT ILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROL LED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROL LED AREA For the purposes of classification, the Security area bet\.veen the OCA detection fence and the PROTECT ED AREA boundary knovvn as the Security OWA-ef Controlled Area (SOGA) in the GGNS Emergency Plan. PROJECT ILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECT ED AREA - An area encompas sed by physical barriers (i.e., the security fence) and to which access is controlled. Page 172 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary.
Basis: This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the SECURITY OWNER CONTROLLED AREA. I The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with 11-S-82-1 Security Contingency Events (ref. 2). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the SECURITY _OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION).\ It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or Page 173 of 270
-::::::- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This i:ncludes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1). Escalation of the emergency classification level would be via IC HS1. Reference(s):
- 1. GGNS Security Plan
- 2.
- 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HA1 Page 174 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bas'es Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by GGNS Security Shift Supervision
- Mode Applicability:
All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy.equipment, take HOSTAGES, and/o.r intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, .or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorisrn-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA). HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER GO.'VTROLLED AREA For the purposes of classification, the Security area betvveen the OGA detection fence and the PROTECTED AREA boundary l<nrn.vn as the Security Owner Controlled Area (SOGA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which, access is controlled. SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. Page 175 of 270
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses the occurrenc*e of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant . equipment. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2). ,\ I Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heighter:,ed state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltecing). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1).
- Reference(s):
- 1. GGNS Security Plan
- 2. 11-S-82-1 Security Contingency Events
- 3. NEI 99-01 HS1 Page 176 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than QBE levels EAL:
HU2.1 Unusual Event Seismic event > QBE as indicated by annunciation of EITHER of the following on SH13P856:
- Containment Operating Basis Earthquake (P856-1A-A3)
- Drywell Operating Basis Earthquake (P856-1A-A5)
Mode Applicability: All Definition(s): None Basis: This IC addresses a seismic event that result~ in accelerations at the plant site greate~ than those specified for an Operating Basis Earthquake (QBE). An earthquake greater than an QBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., perform walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections,, and fully und~rstand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if
- deemed appropriate (e.g., a call to the U.S. Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SAB.1. To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrum~ntation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the QBE alarm. If requested, provide the analyst Page 177 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases with the following GGNS coordinates: 32° O' 27" north latitude, 91° 2' 53" west longitude (ref. 2). Alternatively, near real-time seismic activity can be access*ed via the NEIC website.
Reference(s):
- 1. 05-S-02-Vl-3 Earthquake
- 2. UFSAR 2.1.1 Site Location and Description
)
- 3. NEI 99-01 HU2 Page 178 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment '1 - Emergen.cy Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating ~ondition: *Hazardous event EAL:
HU3.1 Unus uai Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, Sor C. If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1. A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm. Reference( s):
- 1. 05-1-02-Vl-2 Hurricanes, Tornados and Severe Weather
- 2. NEI 99-01 HU3 Page 179 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components tha~ are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Page 180 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Refer to EAL CA6.1 or SAB.1 for internal FLOOD! NG affecting more than one SAFETY SYSTEM train.
Reference(s):
- 1. 05-1-02-Vl-1 Flooding
- 2. NEI 99-01 HU3 Page 181 of 270
{~
~ -::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability: All Definition(s): IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s):
- 1. NEI 99-01 HU3 Page 182 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Note 7: This EAL does not apply to routine traffic impediments such as fog, s*now, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes. an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. *Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C. Reference(s):
- 1. NEI 99-01 HU3 Page 183 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory : 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):
- Report from the field (i.e., visual observation)
- Receipt of multiple (more than 1) fire alarms or indications
- Field verification of a single fire alarm
\
AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table H-1 Fire Areas
- Unit 1 Containment
- Unit 1 Auxiliary Building
- Unit 1 Turbine Building
- Control Building
- Diesel Generator Rooms
- SSW Pump & Valve Rooms Mode Applicability :
All Definition(s) : FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the conditiorfs existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 184 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Leyel Technical Bases Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). lh addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or repo,rt. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SAB.1. The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALi D fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2). Reference(s):
- 1. 05-S-02-V-1 Response to Fires
- 2. 1O-S-03-2 Response to Fires
- 3. NEI 99-01 HU4 Page 185 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) (Note 11) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 11: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the containment or drywell. Table H-1 Fire Areas
- Unit 1 Containment
- Unit 1 Auxiliary Building
- Unit 1 Turbine Building
- Control Building
- Diesel Generator Rooms
- SSW Pump & Valve Rooms Mode Applicability:
All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
- Page 186 of 270
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~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses the mag.nitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) Within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. This EAL is not applicable for the containment or drywell in Modes 1 and 2. The air flow design and TS requirements for operation of Containment Fan Coolers and the drywell cooling system are such that multiple detectors would be expected to alarm for a fire in the containment or drywell. A fire in the containment or drywell in these modes would therefore be classified under EAL HU4.1. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
- Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to ,safety shalr'be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can leaq to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under ,post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. Page 187 of 270
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-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. The 30 minute requirement begins upon receipt of a single VALID fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15 minute requirement beginning with the verification of the fire by field report. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2). Reference(s):
- 1. 05-S-02-V-1 Response to Fires
- 2. 1O-S-03-2 Response to Fires
- 3. UFSAR Appendix 9A Fire Hazard Analysis Report
- 4. NEI 99-01 HU4 Page 188 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: ( All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. i Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
- In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. Reference(s):
- 1. NEI 99-01 HU4 Page 189 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Basis: This IC addresses the magnitude and extent of-FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SAB.1. Reference(s):
- 1. NEI 99-01 HU4 Page 190 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Ba~es Category: H - Hazards and Other Co~ditions Affecting Plant Safety
/ Subcategory: 5 - Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations,. cooldown or shutdown** EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND , Entry into the room or area is prohibited or IMPEDED (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. " T~ble H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode I Control Building 111' SWGR Rms (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A 103) 3 Auxiliary Building 93' RHR B Pump Room (1A 105) -:, 3 Auxiliary Building 93' Corridor (1A 101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203) 3 Auxiliary Building 119' RHRB Pump Room (1A205) 3 Auxiliary Building 119' RCIC Room (1A204) 3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR B Room (1A306, 1A307) 3 Radwaste Building 118'. Radwaste Control Room (OR241) 3 Mode Applicability: 3 - Hot Shutdown
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Page 191 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect aMhe time of the gaseous release. The. emergency classificatio~ is not contingent upon whether entry is actually *necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of pmtective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply:
- The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.
- The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
- The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.
- If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. Page 192 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases .
This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. ' The list of plant rooms or areas with entry-related mode applicability identified specify th.ose rooms or areas that contain equipment which require a manual/local action as.specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1). EAL HAS.1 mode applicability has been limited to the mode limitations of Table H-2 (Mode 3 only). Reference(s):
- 1. Attachment 2 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
- 2. NEI 99-01 HAS Page 193 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel Mode Applicability: All Definition(s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities wiJI assist in responding to these challenges. Transfer of plant control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC HS6. Reference( s):
- 1. 05-1-02-11-1 Shutdown from the Remote Shutdown Panel
- 2. NEI 99-01 HA6 Page 194 of 270
. ~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:
HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1):
- Reactivity (Modes 1 and 2 only)
- RPV water level
- RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.
Mode AppUcability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, arid the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown 16cation(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC FG 1 or CG 1 Page 195 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-1-02-11-1 Shutdown from the Remote Shutdown Panel
- 2. EP FAQ 2015-014 i3. NEI 99-01 HS6 Page 196 of 270
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-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL:
HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.
- Mode Applicability:
All Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during
- and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant
- declaration of an emergency because conditions exist which are believed by the Emergency Director tci fall under the emergency classification level description for an UNUSUAL EVENT.
Reference(s):
- 1. NEI 99-01 HU?
Page 197 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory : 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL:
HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability : All Definition(s) : HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLE D AREA). 01,11/NER GOl\lTROLLED AREA For the purposes of classification, the Security area betvveen the OC/\ detection fence and the PROTECTED /\REA boundary known as the Security 0 1ovner Controlled Area (SOC/\) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -An area encompasse d by physical barriers (i.e., the security fence) and to which access is controlled. SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warni nq capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. Page 198 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT. Reference(s):
- 1. NEI 99-01 HA7 Page 1,99 of 270
~ ~ ~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL:
HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the* protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA). OW.'VER CONTROLLED AREA For the purposes of classification, the Security area between the QC/\ detection fence and the PROTECTED ARE/\ boundary knovvn as the Security Ovvner Controlled Area (SOC/\) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. ' PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. Page 200 of 270
*~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases SECURITY OWNER CONTROLLED AREA - The SOGA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOGA is the area between the SOGA Fence and the PROTECTED AREA Boundary.
SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY. , i. Reference(s):
- 1. NEI 99-01 HS7 Page 201 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL:
HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Mode Applicability: All Definition(s): HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. 0111/NER GOfVTROLLED AREA For the purposes of classification, the Security area bet\.veen the OCA detection fence and the PROTECTED AREA boundary kno\vn as the Security Ovmer Controlled Area (SOGA) in the GGNS Emergency Plan. PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Page 202 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases PROTECTED AREA -An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.
SECURITY OWNER CONTROLLED AREA - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA Boundary. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY. Reference(s):
- 1. NEI 99-01 HG?
Page 203 of 270
-=-==-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failu.re events that warrant emergency classification have been identified in this category. They may pose actual or potentia) threats to plant safety. The events of this category pertain to the following subcategories:
- 1. Loss of ESF AC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4;16 KV ESF buses.
- 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
- 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory. *
- 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures, and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
- 5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core.
The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity. Page 204 of 270
ft ~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases
- 6. RPS Failure J This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, A TWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
- 7. Loss of Communications*
Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.
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Page ios of 270 _
~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory : 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite AC power capability to ESF buses for 15 minutes or longer EAL:
SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 KV buses for~ 15 min. (Note 1) Note 1: The Emergency pi rector should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
- DIV I DG (DG 11)
- DIV II DG (DG 12)
Mode Applicability : 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s) : None Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC ESF buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the ESF buses, whether or not the buses are powered from it. Page 206 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level would be via IC SA 1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 SU1 Page 207. of 270
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-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power
- Initiating Condition: Loss of all but one AC power source to ESF buses for 15 minutes or longer EAL:
SA1.1 Alert AC power capability, Table S-1, to ,DIV I and DIV 11 ESF 4.16 KV buses reduced to a single power source for;;::: 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite
- ESF Transformer 11
- ESF Transformer 12
- ESF Transformer 21 Onsite
- DIV I DG (DG 11)
- DIV II DG (DG 12)
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or
. placing it in the cold shutdown condition, including the ECCS. These are typically systems I classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; Page 208 of 270
ft.
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revi'sion XXX Attachment 1 - Emergency Action Level Technical Bases (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: The HPCS bus (DIV Ill) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.
- A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
- A loss of all offsite power and loss of all ESF emergency power sources (e.g., onsite diesel generators) with a single train of ESF buses being back-fed from the unit main generator.
- A loss of ESF emergency power sources (e.g., orsite diesel generators) with a single train of ESF emergency buses being fed from an offsite pow~r source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1. This EAL is the hot condition equivalent of the cold ¢ondition EAL CU2.1. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power '
- 6. NEI 99-01 SA1 Page 209 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to ESF buses for 15 minutes or longer EAL:
SS1 .1
- Site Area Emergency Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4.16 KV buses for ~ 15 min. (Note 1)
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including t~ose necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink Mitigative strategies using other power sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool coolin'g systems.would be essential subsequent to a station blackout.. In addition, fission product barrier monitoring capabilities may be degraded Page 21 O of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CA2.1. _Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. NEI 99-01 SS1 Page 211 of 270 .
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S -System Malfunction Su~category: 1 - Loss of ESF AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to ESF buses EAL: SG1 .1 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4. 16 KV buses AND EITHER:
- Restoration of at least one ESF 4. 16 KV bus in < 4 hours is not likely (Note 1)
- RPV water level cannot be restored and maintained> -191 in.
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or
- placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): \
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of acciqents which could result in potential offsite exposures: Basis: Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-191 in.) (ref. 6). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling). This IC addresses a prolonged loss of all power sources to AC ESF emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat Page 212 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 .- Emergency Action Level Technical Bases removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the ,emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC ESF emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is a greater likelihood of challenges to multiple fission product barriers. 4 hours is the site-specific SBO coping analysis time (ref. 4). The estimate for restoring at least orie ESF emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
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Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05~1-02-1-4 Loss of AC Power
- 6. 02-S-01-40 EP Technical Bases
- 7. NEI 99-01 SG1 Page 213 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power lnitiaUng Condition: Loss of all ESF AC and vital DC power sources for 15 minutes or longer EAL:
SG1 .2 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV 11 ESF 4.16 KV buses for~ 15 min. (Note 1) AND Indicated voltage is < 105 voe on vital 125 voe buses 11 DA and 11 DB for ~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit ha,~ been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown co,ndjtion, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the conseq'uences of accidents which could result in potential offsite exposures. Basis: Vital DC buses 11 DA and 111 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1. 72 volts/cell. These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC) (ref. 6, 7). This IC addresses a concurrent and prolonged loss of both emergency ESF AC and Vital DC power. A loss of all emergency ESF AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, Page 214 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (HPCS DIV Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to, a station blackout. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency ESF AC and Vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. Reference(s):
- 1. UFSAR Figure 8.1-001 Main One Line Diagram
- 2. UFSAR section 8.1 Electric Power Introduction
- 3. UFSAR section 8.3 Onsite Power
- 4. UFSAR section BA Loss of all AC Power
- 5. 05-1-02-1-4 Loss of AC Power
- 6. Calculation No: EC-01111-14001 Station Division I Battery 1A3 and Division II Battery 183 Discharge Capacity during Extended Loss of AC Power
- 7. UFSAR 8.3.2.1.1 Station DC Power
- 8. NEI 99-01 SG8 Page 215 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document R~vision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S *._ System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:
SS2.1 Site Area Emergency Indicated voltage is < 105 voe on vital 125 voe buses 11 DA and 11 DB for
~ 15 min. (Note 1)
Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequence s of accidents which could result in potential offsite exposures. Basis: Vital DC buses 11 DA and 11 DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1. 72 volts/cell. These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC) (ref. 1, 2). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the prote~tion of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CU4. 1. Page 216 of 270
-===- Entergy Grand' Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. Calculation No: EC-01111-14001 Station Division I Battery 1A3 and Division II Battery 183 Discharge Capacity during Extended Loss of AC Power
- 2. UFSAR 8.3.2.1.1 Station DC Power
- 3. NEI 99-01 SSS Page 217 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. I Table S-2 Safety System Parameters
- Reactor power
- RPV water level
- RPV pressure
- Containment pressure
- Suppression Pool water level
- Suppression Pool temperature Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential'offsite exposures. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 218 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Basis:
This IC addresses the difficulty associated with monitoring normal plant conditions without the . ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed . parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be
,compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via EAL SA3.1. Reference(s):
- 1. UFSAR 7.5 Safety-Related Display Instrumentation
- 2. NEI 99-01 SU2 Page 219 of 270,
Grand Gulf Nuclear Station EAL Basis Document Revision XXX
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Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) AND I Any significant transient is in progress, Table S-3 Note 1 :. The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-2 Safety System Parameters
- Reactor power
- RPV water level
- RPV pressure
- Containment pressure
- Suppression Pool water level
- Suppression Pool temperature Table S-3 Significant Transients
- Reactor scram
- UNPLANNED drop in reactor thermal power > 25%
- Electrical load rejection > 25%
electrical load
- ECCS injection
- Thermal power oscillations > 10%
Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Page 220 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during
- and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room. ( An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accor9ance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, Page 221 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC FS1 or AS1. Reference(s):
- 1. UFSAR 7.3 Engineered Safety Features Systems
- 2. NEI 99-01 SA2 Page 222 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergen.cy Action Level Technical Bases
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Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Offgas Pretreatment radiation monitor high-high alarm Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: The Offgas Pretreatment monitors radioactivity in the Offgas system downstream of the Offgas condenser. The monitor detects the radiation level that is attributable to t~e fission gases producyd in the reactor and transported with steam through the turbine to the condenser. The Hi-Hi alarm, if alarming, indicates that the radioactivity present at the recombiner effluent discharge is at or above the Technical Specification 3. 7.5 limit of 380 millicuries per second of Noble. Gases. (ref. 1) This IC addresses a reactor coolarit activity value that exceeds an allowable limit specified in Technical Specifications. This condition is* a precursor to a more significant event and represents a potential degrad,ation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs. !Gn the event that the Offgas Pretreatment Radiation Monitor High-High Alarm is_out of service, the use of offgass flowrates and Offgas Pretreatment Radiation monitor readings is a viable contigiency contingency action to classify the EALiG. See chart in 04-1-02-1 H 13-P601-19A-D7 Alarm RespQnse Instruction for OG PRE-TREAT RAD HI_HI alarm. Reference(s):
- 1. Alarm Response Instruction 04-1-02-1 H13-P601-19A-D7
- 2. UFSAR 11.5 Process and Effluent Radiological Monitoring and Sampling Systems
- 3. Technical Specification 3.7.5 Main Condenser Offgas
- 4. 05-1-02-11-2 Offgas Activity High
- 5. NEI 99-01 SU3 Page 223 of 270
~~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 ~ RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:.
SU4.2 Unusual Event Coolant activity> 0.2 µCi/gm dose equivalent 1-131 for> 48 hours OR Coolant activity> 4.0 µCi/gm dose equivalent 1-131 instantaneous Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA 1 or the Recognition Category A ICs. Reference(s):
- 1. Technical Specification 83.4.8, RCS Specific Activity bases
- 2. UFSAR Section 15.6.4 _Steam System Piping Break Outside Containment
- 3. NEI 99-01 SU3 Page 224 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:
SU5.1 Unusual Event RCS unidentified or pressure boundary leakage> 1O gpm for~ 15 min. (Note 1) OR RCS identified leakage> 25 gpm for~ 15 min. (Note 1) OR Leakage from the RCS to a location outside containment > 25 gpm for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a collecting sump; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. Unidentified leakage is all leakage into the drywell that is not identified leakage (ref. 2, 3). Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2', 3). This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS , Page 225 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve doe? not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold du"ration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of Recognition Category A or F. Reference(s):
- 1. UFSAR Section 5.2.5, Detection of Leakage Through Reactor Coolant Pressure Boundary
- 2. Technical Specification Definitions Section 1.1
- 3. Technical Specification 3.4.5
- 2. NEI 99-01 SU4 Page 226 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:
SU6.1 Unusual Event . An automatic scram did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in shutting down the reactor as indicated by reactor power::; 5% (APRM downscale) (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. The first condition of this EAL identifies the need to cease critical reactor operations by . actuation of the_automatic Reactor Protection System (RPS) scram function. A reactor scra.m is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1). i \ A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%. 1 For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2). Page 227 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX 1 1 Attachment 1 - Emergency Action Level Technical Bases Following any automatic RPS scram signal, operating procedures (e.g., EP-2) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts
*a11 control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event (ref. 3).
I Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor pow~r remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to s 5% is not considered a successful automatic scram. If automatic initiation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI/RPT is an acceptable means of , establishing reactor shutdown conditions relative to the EA.L threshold in the absence of any required subsequent manual scram actions. In the event that the operator identifies a reactor scram is IMMINENT and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is r required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50. 72 should be considered for the transient event. Following1 the failure of an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. Page 228 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. J The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful i,n shutting down the reactor, then the emergency classification level will escalate to an Alert via EAL SA6.1. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.
Should a reactor scram signal be generated as a result of plant work {e.g., RPS setpoint testing), the following classification guidance should be applied.
- If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to .
automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
*
- If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.
Reference(s):
- 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation
- 2. 05""S-01-EP-2A A TWS RPV Control
- 3. 05-S-01-EP-2 RPV Control
- 4. NEI 99-01 SUS Page 229 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:
SU6.2 Unusual Event A manual scram did not shut down the reactor as indicated by reactor power > 5% after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in shutting down the reactor as indicated by reactor power :s; 5% (APRM downscale) (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown co,,ndition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor cc;mtrol consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power~ 5%) (ref. 1). Page 230 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale ~etpoint of 5%.
For the purposes of emerge*ncy classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2). Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. Successful automatic or manual initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power. associated with the SAFETY SYSTEM design (~ 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1. Following the failure of an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch. Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). Thi's action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room; or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is consipered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram wifl vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual ac~ions taken at the reactor control consoles Page 231 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA6 or FA 1, an Unusual Event declaration is appropriate for this event.
Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
- If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
- If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.
Reference(s):
- 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation
- 2. 05-S-01-EP-SA A TWS RPV Control
- 3. 05-S-01-EP-2 RPV Control
- 4. NEI 99-0j SU5 Page 232 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6- RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1 Alert An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% AND Manual scram actions taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) are not successful in shutting down the reactor as indicated by reactor power> 5% (Note 8) I Note 8: A manual scr~m action is any operator action, ~r set of actions, which causes the* control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;
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(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action Page -233 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases.
taken away from the reactor control consoles since this event entails a significant failure of the
. RPS.
This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by subsequent manual scram actions that fail to shut down the reactor to an J extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (> 5%). For the purposes of emergency classification,, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2). For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to or below 5% is not considered a successful automatic scram. If automatic actuation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor
, scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g.,
locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, esca'lation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. Page 234 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation
'2. 05-S-01-EP-2A A TWS RPV Control
- 3. 05-S-01-EP-2 RPV Control
- 4. NEI 99-01 SAS Page 235 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: , Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL:
SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor as indicated by reactor power
>5%
AND All actions to shut down the reactor are not successful as. indicated by reactor power> 5% AND EITHER: RPV water level cannot be restored and maintained> -191 in. OR Heat Capacity Temperature Limit (HCTL) exceeded (EP Figure 1) Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following d~sign basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. Page 236 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases This EAL addresses the following:
- Any automatic reactor scram signal followed by subsequent manual scram actions that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1 ), and
- Indications that either core cooling is extremely challenged or heat removal is extremely challeng~d.
Reactor shutdown achieved by use of control rod insertion methods in EP.:.2A step Q-1. are also credited as a successful shutdown provided reactor power can be reduced to or below the APRM downscale trip setpbint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1) The combination of failure of both front line and backup protection systems to .function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers. Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 1). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence (ref 2). The Heat Capacity Temperature Limit (HCTL, EP Figure 1) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppression pool temperature. The HCTL is a Junction of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step of section SPT in EP-3, Containment Control, is reached (ref. 3). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. In some instances, the emergency classification resulting from this EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this EAL ensures the tim'ely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. Escalation of the emergency classification level would be via IC AG1 or FG1. Page 237 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Reference(s):
- 1. 05-S-01-EP-2A, ATWS RPV Control
- 2. 05-S-01-EP-5, RPV Flooding
- 3. 05-S-01-EP-3, Containment Control
- 4. NEI 99-01 SS5 Page 238 of 270
~=-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communication*s Initiating Condition: Loss of all onsite or offsite communications. capabilities i
EAL: SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and lpcal agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods State/ Syste~ Onsite NRC Local Station Radio System x ., GGNS Plant Phone System x Public Address System x Emergency Notification System (ENS) x Commercial Telephone System x x Satellite Phones x x INFGRM NetifieatieR System x Operational Hotline x Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Page 239 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachm~nt 1 - Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge-to plant or personnel safety, this event warrants prompt notifications to State* and local agencies and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriff's Department, Louisiana Department of Environmental Quality, Tensas Parish Sheriff's Office, and the Louisiana Governor's Office of Homeland S~curity and _Emergency Preparedness. The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the hot condition equivalent of the cold condition EAL CU5.1. Reference(s):
- 1. GGNS Emergency Plan Section 7.5, Communications Systems
- 2. 04-S-01-R61-1 Plant Communications
- 3. NEI 99-01 SU6 Page 240 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
SA8.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
- Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operatin_g mode
- Event damage has* resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table S-5 Hazardous Events
- Seismic event (earthquake)
- Internal or external FLOODING event
- High winds or tornado strike
- FIRE
- EXPLOSION
- Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Page 241 of 270
-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpr~ssurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are ,observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern reg~rding the operability or reliability of the affected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either i'ndications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the* potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
- Page 242 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 1 - Emergency Action Level Technical Bases ( . . Indications of degraded performance addresses damage to a SAFETY SYSTEM train th.at is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC FS1 or AS1. This EAL is the hot condition equivalent of the cold condition/EAL GA6.1. Reference(s):
- 1. EP FAQ 2016-002
- 2. NEI 99-01 SA9 Page 243 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases I
Background
NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be 'plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified". should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope. Page 244 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 &- H-2 Bases GGNS Table A-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown: IOI / 501 ACTIONS IOI 03-1-01-2 Power Operations LOWER Power by reducing Recirculation flow until 62.2% core flow (70 MCR 1 mlbm/hr) is reached. INSERT Control Rods per Control Rod Movement Sequence. MCR 1 TECH SPEC TRIGGER (SR 3.3.2.1.2, SR 3.3.2.1.4) MCR 1 IF Reactor power has been reduced below the HPSP OR the LPSP, THEN PERFORM one of the following: Required Surveillances or enter LCO for TS 3.3.2.1 ' CHECK OPEN the following valves on 1H13-P870-6C: MCR
- a. N11-F029A, HP TUR8 EXTR To MSRA 1ST STG RHT
- b. N11-F0298, HP TUR8 EXTR To MSR 8 1ST STG RHTIF N11-F029A OR N11-F0298 are NOT open; THEN RETURN MSR 1ST Stage Reheaters to service per SOI 04-1-01-N11-1.
CHECK OPEN the following valves on panel 1H13-P870-6C: MCR 1
- a. N36-F01 OA, EXTR STM SPLY TO FW HTR 5A
- b. N36-F0108, EXTR STM SPLY TO FW HTR 58
- c. N36-F011A, EXTR STM SPLY TO FW HTR 6A
- d. N36-F0118, EXTR STM SPLY TO FW HTR 68 TAKE handswitches for the following valves to OPEN position on panel 1H13-P870-6C:
- a. N36-F013A, FW HTR 5A EXTR STM 8TV
- b. N36-F0138, FW HTR 58 EXTR STM 8TV
- c. N36-F012A, FW HTR 6A EXTR STM 8TV
- d. N36-F0128, FW HTR 68 EXTR STM 8TV NOTIFY the following of the power reduction: MCR 1
- Load Dispatcher (Woodlands)
* *Duty Manager (IF u.nexpected power reduction) * (SMEPA)(1-601-261-2318 OR 1-601-261-2313) * * (SMEPA) Site Representative
- Radwaste
- Radiation Protection
- Chemistry
* *NRC Resident Inspector Page 245 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / 501 ACTIONS
- LOCATION MODE NOTES
- These notifications Must be made by Shift Manager IF LP Turbine inlet temperature is >491°F, and N11-F028A and N11- TURB BLDG 1 Not F028B are open, THEN SIMULTANEOUSLY THROTTLE the following ELEV 133 Required valves on 1H22-P177 to CONTROL LP Inlet Temperatures within a AREA4 band of 470° F to 490° F while monitoring LP Turbine Inlet differential ROOM temperatures within 30° F (comparing A side to B side). 1T325
. N11-F028A . N11-F028B *I IF LP Turbine inlet temperature is >491 °F, and N11-F028A and N11-F028B are closed, THEN SLOWLY, SIMULTANEOUSLY LOWER MSR-A/B HTG STM FEED CONT manual setpoint to CONTROL LP Inlet Temperatures within a band of 470° F to 490° F while monitoring LP Turbine Inlet differential temperatures within 30° F (comparing A side to B side).
LOWER Reactor power by INSERTING control rods to specified MCR 1 Control Rod in-sequence position per 17-S-02-400. At approximately 48% Reactor power, PERFORM the following on MCR 1 panel 1H13-P601. VERIFY the following valves Open:
- B21-F033 INBD MSL DR SOL TO MN CNDSR
- B21-F069 OTBD MSL DR SOL TO MN CNDSR
- OPEN B21-F016 At approximately 50% Reactor Power, PERFORM the following:
SHUTDOWN 1 Reactor Feed Pump per 501 04-1-01-N21-1. VERIFY RFPT B is operating normally on master controller. MCR 1 RAISE FW MASTER LVL CONT setpoint to approximately 39" MCR 1 TRANSFER the RFPT A SP CONT to MAN. MCR 1 SLOWLY LOWER speed of RFPT A USING RFPT A SP CONT by MCR 1 DEPRESSING the OUT D pushbutton. OBSERVE speed of RFPT B raises to maintain RPV water level, OR control it manually FURTHER REDUCE speed of RFPT A using RFPT A SP CONT MCR 1 in MAN until it reaches low speed stop. TRANSFER speed control of RFPT A to SPEED AUTO by MCR 1 DEPRESSING the OBSERVE the FW AUTO pushbutton extinguishes AND the SPEED AUTO, RAISE, AND LOWER pushbuttons backlight. FURTHER REDUCE RFPT A speed using RFPT A LOWER MCR 1 pushbutton. WHEN RFPT A speed reaches 1100 rpm, THEN TRIP RFPT A by MCR 1 DEPRESSING the RFPT A MAN TRIP pushbutton CHECK F014A, RFP A DISCH VLV starts to close. I MCR 1 Page 246 of 270
-====- Entergy Grand Gulf Nuclear Station EAL Basis Oocument Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES REOPEN F014A, RFP A DISCH VLV WHEN RFPT A coasts down to zero speed, THEN RESET turning gear MGR by pressing A TURN GEAR OPER RESET pushbutton.
OBS.ERVE turning gear engages automatically, unless RFPT A is rolling on min flow. IF turning gear fails to engage, THEN MANUALLY ENGAGE the TURB BLDG Not turning gear locally by PRESSING DOWN the manual engaging lever. ELEV 133 Required AREA3 ROOM 1T307, 1T309 CHECK OPEN/OPEN the following Drain valves on 1H22-P175: N/A N/A These 1N11-F019A, RFPT A HP IN DRVLV steps are not 1N11-F023A, RFPT A HP IN DR VLV required 1N11-F018A, RFPT A IP IN DR VLV to be 1N11-F021A, RFPT A IP IN DR VLV performed 1N11-F042A, RFPT A IP IN DRVLV to Shut down and 1N33-F021A, RFPT A ABOVE SEAT DR Cool down 1N33-F022A, RFPT A ABOVE SEAT DR the plant. 1N33-F023A, RFPT A BELOW SEAT DR 1N33-F024A, RFPT A BELOW SEAT DR RETURN FW MASTER LVL CONT setpoint to approximately 36" MGR IF desired, RESET RFPT A trip using the RFPT A TRIP RESET MGR pushbutton SHUTDOWN 1 Circul_ating Wtr Pump per SOI 04~1-01-N71-1 CHECK that CTCS balls are collected AND system shut down. DEPRESS the BALL CATCH FLAP CATCH pushbutton on P001A (B) Turb Bldg 1 Not MIMIC AND OBSERVE the flap rotates to the CATCH position. 113' Area 4 Required (1T203) OBSERVE ball collection starts by a rising number of balls in ball
- Turb Bldg 1 Not collector tank. 113' Area 4 R~quired (1T203)
After 10 minutes STOP Ball Recirculation pump by DEPRESSING Turb Bldg 1 Not RECIRC PUMP OFF pushbutton on P001A(B) MIMIC 113' Area 4 Required (1T203) CLOSE Pump Discharge Valve F323A(B). Turb Bldg 1 Not 113' Area 4 Required (1T203) PLACE Screens #1 AND #2 in BACKWASH position by DEPRESSING Turb Bldg 1 Not Page 24 7 of 270
-~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES SCREEN BACKWASH pushbutton on P001A (B) MIMIC AND 113' Area 4 Required OBSERVE screens rotate to BACKWASH position. (1T203)
PRESS the CIRC WTR PMP A(B) STOP pushbutton on 1H13-P680. MCR CHE.CK that F002A(B) Circulating Water Pump Discharge valve closes MCR on 1H 13-P680 ENSURE that A(B) Circulating Water pump has shut down USING MCR 1 pump indication light on 1H13-P680 WHEN its discharge valve is CLOSED. OPEN OR CHECK OPEN F001 USING CIRC WTR LOOP NB XTIE MCR hand switch on 1H13-P870. CLOSE OR CHECK CLOSED F040A (B) Acid Feed Valve. N/A N/A Not required to be performed to Shut down and Cool down the plant. CLOSE OR CHECK CLOSED LV-F513 A(B), Slowdown valve MCR OPEN F039A(B), CIRC WTR PUMP A(B) COLUMN VENT N/A N/A Not required to be performed to Shut down and Cool down the plant. ENSURE Condenser vacuum is maintained> 23.8" Hg* MCR 1 SHUTDOWN one Heater Drain Pump per SOI 04-1-01-N23-1 JOG CLOSED N23-F051A(S), HTR DR PMP A(B) DISCH VLV on 1H13-P680 for desired pump. STOP HTR DR PMP A(B) on 1H13-P680. WHEN Reactor power has been reduced < 40%, SHUTDOWN 2nd Heater Drain Pmp per SOI 04-1-01-N23-1 Before securing second Heater Drain Pump, PLACE N23-LK-R053, MCR 1 HTR DR TK DR, in Manual AND Slowly REDUCE output to 0%. ENSURE Heater Drain Tank level is maintained by Dump Valves MCR N23-LV-F518A-E JOG CLOSED N23-F051 B(A) HTR DR PMP B(A) DISCH VLV on MCR 1 1H13-P680 for second pump. Page 248 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / 501 ACTIONS LOCATION MODE NOTES STOP Heater Drain Pump HTR DR PMP B(A) on 1H13-P680. MCR 1 WHEN BOTH Heater Drain Pumps are shutdown, TURB BLDG Not THEN CLOSE N23-F054, HTR DR PMP COMMON DISCH VLV on ELEV 133 Required 1H22 P175 AREA6 ROOM 1T327 IOI 03-1 ..01-2 Continued SHIFT the Reactor Recirculation ~ump(s) to slow speed as follows: MCR 1 INSERT Control Rods until Load Line is between 50 AND 65% VERIFY Control Rods are in sequence of the Control Rod Pattern Controller. BEFORE entering Controlled Entry Region of Figure 3, PERFORM the MCR 1 folloVl(ing WHEN TS 3.3.1.1, Action J.1 is in effect: VERIFY Fraction of Core Boiling Boundary (FCBB) is s 1.0 per 06 RE-1J11-V-0002. IMPLEMENT TS 3.3.1.1, Action J.2, within 12 hours of entry AND J3 within 96 days. IF any APRM gain is out of tolerance, THEN ADJUST gain per 06-RE- MCR 1C51-W-0001 prior to downshift of Recirculation Pumps. CLOSE Both Recirculation A AND B Flow Control Valves (FCV's) to MCR Min Ed position using RECIRC A(B) FLO CONT on 1H13-P680 TRANSFER Both Reactor Recirculation Pumps to slow speed per MCR *1 SOI 04-1-01-833-1 CONTINUE Reactor Power reduction to 25 - 30% by insertion of MCR 1 Control Rods SHUTDOWN Hydrogen Water Chemistry Injection per 501 04-1-01-P73-1. At H13-P845, momentarily DEPRESS HWC SHUTDOWN pushbutton MCR 1 AND OBSERVE the following: HWC SHUTDOWN pushbutton 1P73-M602 Will be flashing as H2 AND 02 flows ramp down to 0. 02 isolation valv.es Will Close WHEN 02 levels remain at normal levels with no 02 injection for at least 5 minutes. HWC SHUTDOWN pushbutton Will be in solid WHEN all control valves AND isolation valves are fully Closed. HWC RUNNING pushbutton extinguishes. CLOS.E P73-F107, H2 lnj Sply Line Man Line Shutoff valve. N/A N/A Not required After 02 valves F515 AND F512 (as indicated by white dots on red cap to be being perpendicular to pipe) have Closed, CLOSE OR CHECK ; CLOSED Both F207 AND F208, 02 Rack Sply lsol to OG Preheater performed to Shut A(B) .. Page 249 of 270
~~ -====- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES CLOSE 1P73-F209, 02 injection to Condensate pumps down and Cool down the plant.
IF Drywell entry is scheduled, WHEN Reactor Power has been reduced N/A N/A Not to less than 30%, THEN PERFORM the following: required PERFORM the following for 1021-K607, DRWL PERS HATCH ARM: to be performed DIRECT l&C to CONNECT Canon plug to the plug labeled "ALARM" to Shut AND "J3" at the back of 1021 K607. down and Cool down PLACE Function Selector switch on front of 1021 K607 (DRWL PERS the plant. HATCH ARM) to OPERATE position. PERFORM EPI 04-1-03-021-1 for 1021 K607. IOI 03-1-01-2 Continued REMOVE Both Second Stage MSR Reheaters from service per SOI 04-1-01-N11-1. OBSERVE PDS Computer Points N11 N044A,B,C AND N11 N045A,B, MCR C to monitor LP Turbine Inlet Temperature DT during removal of Second Stage Reheaters from service. ENSURE Both MSR HTG STM FEED CONT are in MANUAL on MCR 1H13-P680. CLOSE the following MSR 2ND STG HTG STM valves on 1H13-P680: MCR N11-F304C N11-F304D SIMULTANEOUSLY CLOSE the following MSR 2ND STG HTG STM MCR 1 valves on 1H 13-P680: N11-F304A N11-F304B LOWER the manual outputs on Both MSR HTG STM FEED CONT to MCR minimum on 1H13-P680 to close the temperature control valves. CLOSE the following MSR SUPPLY VLVS valves on 1H22-P177. TURB BLDG 1 Not N11- F028A ELEV 133 Required N11- F028B AREA4 ROOM 1T325 VERIFY the following valve lineup on local panels: N/A N/A Not N35-F015A Closed, HS-M003A required N35-F015B Closed, HS-M003B to be performed N35-F018A Closed, HS-M007A to Shut N35-F018B Closed, HS-M007B down and Cool down the plant.- Page 250 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES IF Feedwater Heater 6A/B are being supplied from extraction steam TURB BLDG Not (i.e., IF 1N36-F01 OA/B AND 1N36-F011A/B on 1H13-P870 are open), ELEV 133 Required THEN CLOSE the following valves on 1H22-P177: AREA4 N35-F008A, 2ND STG RHTR DR TK A TO HTR 6A ROOM 1T325 N35-F008B, 2ND STG RHTR DR TK B TO HTR 68 REMOVE Both First Stage MSR Reheaters from service .. per SOI 04-1-01-N11-1.
OPEN the following valves on 1H13-P870: MCR 1 N11-F005A, MSR 1ST STG RHT RO BYP DR VLVS N11-F005B, MSR 1ST STG RHT RO BV:P DR VLVS SIMULTANEOUSLY CLOSE the following valves on 1H13-P870: MCR 1 N11-F029A, HP TURB EXTR TO MSRA 1 N11-F029B, HP TURB EXTR TO MSR B CLOSE the following valves by taking its respective handswitch to MCR TEST: N11-F003A, MSR A 1ST STG RHT EXTR STM BTV (1H13..: P870) N11-F003B, MSR B 1ST STG RHT EXTR STM BTV (1H13-P870) REMOVE Condensate Precoat filters from service per Not 501 04-1-01-N22-1, IF in service. required to be performed to Shut down and
- Cool down the plant.
OPEN the following BSCV UPSTRM DR VLV's: MCR 1
- a. N33-F300A
- b. N33-F300B
- c. N33-F300C At approximately 23 - 26 % Reactor Power, RAISE the SPEED MCR 1 DEMAND setpoint to approximately 35%, as monitored on PDS Computer point N32K246, by DEPRESSING the SP DEMAND RAISE AND REL pushbuttons.
SIMULTANEOUSLY DEPRESS LOAD REF OFF AND REL MCR 1 pushbuttons on 1H13-P680-9C to turn off load demand Control AND VERIFY OFF light is illuminated. LOWER load by DEPRESSING SpEED DEMAND LOWER AND REL MCR 1 pushbutton. (Expected value 150 :.._ 175 MWe) OBTAIN Shift Manager permission for Manual Scram MCR 1/2
\ Page 251 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 ...:. Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES NOTIFY the following that Main Generator is b1eing disconnected from MCR 1/2 the grid:
- Entergy Load Dispatcher (Woodlands)
* (SMEPA) 1-601-261-2318 OR 1-601-261-2313)
- Entergy Mississippi Dispatcher
- Duty Manager VERIFY Switchyard lineup is acceptable for trip of J5228 AND J5232 MCR 1/2 INSERT IRMs MCR 1/2 NOTIFY the following personnel/departments that a manual scram is MCR 1/2 being initiated:
- Radwaste
- Chemistry
- Radiation Protection -
ANNOUNCE over plant pager that manual Scram is being initiated. TAKE initial temperature data per Attachment Ill, Data Sheet I of IOI MCR 1/2 03-1-01-3 prior to scram Manually SCRAM the Reactor using the MANU.AL SCRAM MCR 1/2/3 pushbuttons.
- a. VERIFY all Control Rods are fully inserted.
- b. VERIFY Reactor Power is de~reasing.
- c. IF Pressure Control System is maintaining reactor pressure greater than 850 psig, THEN PLACE Reactor Mode switch to SHUTDOWN.
- d. VERIFY Reactor Recirculation pumps are running 1n slow speed.
ENSURE Main Turbine and Generator trip. (Reverse power 15 MCR 3 seconds time delay, 5 seconds time delay IF turbine has already tripped.).
- a. VERIFY the Generator Output Breakers open.
- b. VERIFY the Turbine Stop and Control Valves close.
WHEN reactor water level Can be restored AND maintained above 11.4 MCR 3 inches, THEN PERFORM the following to prevent Reactor water level from reaching Level 9 RFPT trip setpoint (58 in.): IF Reactor pressure is dropping rapidly~ THEN SELECT SPEED AUTO OR MANUAL on the running Reactor Feed Pump AND LOWER Reactor Feed Pump discharge pressure to MAINTAIN Reactor level below 58 inches: TRANSFER Feedwater Control to Start-Up Level Control per SOI 04-1-01-N21-1. (Attachment VII of SOI 04-1-01-N21-:1 May be (
- used.)
ENSURE;: Scram Discharge Volume Vent AND Drain valves closed MCR 3 Page 252 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS IOI 03-1-01-4 SCRAM Recovery INSERT all SRM's AND VERIFY response on SRM recorders. MCR 3 SWITCH IRM/APRM LVL recorders to IRM AND VERIFY neutron MCR 3 monitoring established on IRM's IF scram signal Can be cleared AND Reactor level AND pressure are MCR 3 stable, THEN RESET scram AND RETURN CRD System to normal as follows:
BYPASS Scram Instrument Volume* High Level signal by PLACING CRD DISCH VOL HI TRIP BYP switches RPS Div 1, 2, 3, 4 to* BYPASS. . RESET scram by PLACING SCRAM RESET handswitches RPS Div 1, 2, 3, 4 to RESET. VERIFY all CRPs settle into Position '00'. IF any Control Rod is NOT at the 'QO' position, THEN PERFORM one notch insert to attempt to force the rod to settle into the '00' position. WHEN "CRD DISCH VOL WTR LVL HI TRIP" annunciator is clear, THEN RETURN CRD DISCH VOL HI TRIP BYP switches to NORMAL. VERIFY that the HCU scram accumulators have been recharged by OBSERVING the ACCUM FAULT indicating lights on 1H13-P680 are out. THROTTLE G33-F102 to raise bottom head drain flowAND limit MCR 3 Bottom Head Drain Line Heatup/Cooldown to < 100°F/HR. Bottom head drain flow greater than 250 gpm May be required. IF Reactor water level is high, THEN REJECT water to Main Condenser per SOI 04-1-01-G3.3-1 to MAINTAIN level band. PLACE NS SSS OTBD MOV TEST handswitch on 1H 13-P601-19B to MCR 3 the TEST position. VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 the TEST position. VERIFY th.at "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13- P601- '19A-G3) Alarms. ADJUST F033, RWCU SYS BLWDN F/0 CONTVLV is - 10% Open. MCR 3 OPEN OR CHECK OPEN the following valves: MCR 3 F028 RWCU BLWDN CTMT INBD ISOL 1H13-P680 F034, RWCU BLWDN CTMT OTBD ISOL 1H13-P680 IF rejecting to main condenser, OPEN OR CHECK OPEN in the MCR 3 following order: F046 RWCU BLWDN TO MN CNDSR 1H13-P680 Page 253 of 270
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~===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 &_ H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES F041 RWCU BLWDN TO MN CNDSR BYP 1H13-P680 F235 RWCU BLWDN TO MN CNDSR 1H13-P870-3C F234 RWCU BLWDN TO MN CNDSR 1H13-P870-9C IF desired, while rejecting during depressurized OR low pressure MCR 3 conditions, F031, RWCU BLWDN ORF BYP VLV May be Open to allow maximum flow Begin rejecting by SLOWLY OPENING F033, RWCU SYS BLWDN MCR 3 FLO CONT valve, AND IF necessary THROTTLING CLOSED F042 OBSERVE FI-R602, RWCU BLWDN FLO indicator on 1H13-P680 MCR 3 MONITOR reactor water level, blowdown flow AND area/room MCR 3 temperature indication while reject is in progress.
ENSURE Bypass valves are maintaining Reactor pressure MCR 3 IF proceeding to Cold Shutdown, THEN PERFORM Cooldown per MCR 3 Attachment II of IOI 03-1-01-3 concurrent with remaining steps of this attachment. DEPRESS the MHC START DVC "LOWER" pushbutton on 1H13- MCR. 3 P680-9C to reduce the MHC START DVC to Zero. CONFIRM the following Bleeder Trip valves are Closed: MCR 3
- a. N36-F013A, FW HTR SA EXTR STM BTV
- b. N36-F013B, FW HTR 58 EXTR STM BTV
- c. N36-F012A, FW HTR 6A EXTR STM BTV
- d. N36-F012B, FW HTR 68 EXTR STM BTV
- e. N11-F003A, MSRA 1ST STG RHT EXTR STM BTV
- f. N11-F003B, MSR B 1ST STG RHT EXTR STM BTV ENSURE Seal Steam Pressure AND Reactor Feed Pump operation MCR 3 maintained by main steam CLOSE the following valves as soon as possible following Turbine trip N/A N/A Not at Gas Rack 1N44D001-N to isolate Hydrogen Pressure Regulators required N44-PCV-F505 AND F506: to be
- a. N44-FA20 performed
- b. N44-FA21 to Shut down and Cool down the plant.
OBSERVE the following actions occur: MCR 3 Aux Field amps AND generator output voltage indicate 0. Generator field Primary breaker Will trip on a generator/transformer lockout condition (including Water Gire reverse power) AND the TVR feeder switch Will open IF a lockout was Pump can NOT initiated WHEN the Turbine speed drops to -1620 rpm. be verified TURB AUX OIL PMPS A, B OR C starts at about 1335 rpm. running by AUX PW CIRC PUMP starts at about 815 rpm. (Locally) computer Page 254 of 270
*-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES I
TURB SHAFT LIFT OIL PMP starts at about 510 rpm. point in TURB GEAR OIL VLVs N34-FE01/FE02 open at about 210 rpm. the MCR. THROTTLE P43-F053 to maintain Main Turbine Lube Oil temp N/A N/A Not between 90-119DF. required I to be performed to Shut down and Cool down the plant. WHEN fast speed trend recording is no longer necessary AND vessel MCR 3 level is greater than 11.4" AND vessel pressure is less than 1064. 7 psig., THEN PERFORM the following: DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON 821-R623A on 1H13-P601-208. DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON 821-R6238 on 1H13-P601-178. OPEN the Generator motor operated air break GEN DISC J5230. MCR 3 PLACE Red Tag on the Control Room handswitch for J5230 in open position. (This step May be performed after step 9.30.3) AFTER GEN DISC J5230 is opened, THEN PERFORM the following: MCR 3 IF tripped, THEN RESET the following Generator reverse power relays by PRESSING the relay reset rod upwards:
- a. 432/G12 (1 N41-M752)
- b. 432/UT11 (1 N41-M756)
AFTER Generator reverse power relays are reset, THEN RESET the following Generator Lockout relays, IF tripped:
- a. 486-1/G12 (1N41-M769)
- b. 486-2/G12 (1 N41-M770)
- c. 786-1/UT11 (1N41-M759)
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- d. 786-2/UH11 (1N41-M760)
AFTER all Generator Lockout relays are reset AND "GEN UNIT TRIP" MCR 3 annunciator clears on 1H13-P680-9A-A8, THEN OBTAIN Entergy Mississippi dispatcher's permission AND PERFORM the following to close breakers J5228 AND J5232 from 1H13-P680 panel: PLACE SYNC CONT BRKR J5228 switch to ON position. CLOSE 500 KV BRKR J5228. PLACE SYNC CONT BRKR J5228 switch to OFF position PLACE SYNC CONT BRKR J5232 switch to ON position. CLOSE 500 KV BRKR J5232. PLACE SYNC CONT BRKR J5232 switch to OFF position. IF all Generator Lockout relays Will NOT reset, THEN PERFORM the MCR 3 Page 255 of 270
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~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES following:
CONTACT Electrical Maintenance to investigate reason any other Generator relays other than reverse power May have tripped. REQUEST Entergy Mississippi dispatcher to open disconnects to de-energize breaker(s) J5228 AND J5232. DEPRESS EHC SP DEMAND LOWER AND REL pushbutton on 1H13- MCR 3 P680-9C to reduce SP DEMAND indicator to O percent. WAIT for SP LTD meter to decrease to O percent. At each Main Transformer Control Cabinet (Phase A, Phase B, AND Outside at 3 Not Phase C), MN XFMRs Required VERIFY lead cooler group fans are OFF SECURE the following steam loads to limit plant cooldown: SJAE per SOI 04-1-01-N62-1 CLOSE Recombiner Drain Valves N64-F264 AND F265 (N64-F268 93' OG 3 Not AND F269) Pre heater Required A/B Rooms 1T1091T110 CLOSE N64-F007A(B) Preheater Inlet Drain using handswitch on N64- 113' Turb 3 Not P001. Area 1 Required 1T202 OPEN RECOMBINER AIR PURGE A(B) Manual Valve 1N64-F004A(B) 93' OG 3 Not Train A(B) Purge Air Sply Sol Byp for the corresponding recombiner Preheater Required train to ESTABLISH a purge flow of approximately 60 scfm through the A/B Rooms recombiner train. 1T109 1T110 CLOSE N62-F003A(B) CNDSR AIR TO 1 STG SJAE A(B) locally at 133' Turb 3 Not 1H22 P176 Area 1/4 Required OBSERVE that F003A(B) CNDSR AIR TO 1 STG SJAE A(B) indicates 1T305, Closed before continuing to the next step. 1T324 DEPRESS N62-F003A(B) SJAE A(B) 1ST STG SUCT VLV CLOSE MCR 3 pushbutton on 1H13-P680 [1 OC]. CHECK the indication on 1H13-P680 and the following valves Close: MCR 3 SJAE A(B) 1ST STG STM INL VLV, N62-F024A(B) SJAE ICNDSR DR VLV, N62-F011A(B) SJAE A(B) 2ND STG SUCT VLV, N62-F006A(B) SJAE A(B) MN STM SPLY VLV, N62-F001A(B) SJAE A(B) EXH VLV, N62-F012A(B) SJAE A(B) SEP DR VLV, N62-F002A(B) REDUCE setpoint of N62-PIC-R01 OA(B) to zero O psi 113' Turb 3 Not Area 1 Required 1T202 Page 256 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX
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Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES ENSURE OPEN following handswitches on 1H22-P176: 133' Turb 3 Not N62-F004A the COND AIR TO 1 STG SJAE A Area 1/4 Required N62-F004B, the COND AIR TO 1 STG SJAE B 1T305, 1T324 ENSURE OPEN N62-F034 A, B, C, DISCH PIPE ORN VLV for draining 113' Turb A 3 Not discharge piping. MVP Area Required 1T218 WHEN discharge piping has drained, 113' Turb A . 3 Not THEN CLOSE N62-F034 A, B, C, DISCH PIPE ORN VLV. MVP Area Required 1T218 OPEN N62-F014 MECH VAC PUMPS COM SUCTVLV, at 1H22-P176. 133' Turb 3 Not Area 1/4 Required I 1T305, 1T324 , ENSURE proper mechanical vacuum pump oil level (>50%), 113' Turb A 3 Not THEN Prelube with manual oiler as follows: MVP Area Required ENGAGE manual oiler pump handle 1T218 I ROTATE for a minimum of 60 seconds. DEPRESS each plunger 5 times CHECK oil flow visible from each oil return- line. CLOSE P44-F348 A(B,C) MECH VAC PMP COOLER DRAIN. 113' Turb A 3 Not OPEN P44-F109 A(B,C) MECH VAC PMP PSW INL ISOL. MVP Area Required OPEN P44-F344 A(B,C) MECH VAC PMP PSW DISCH ISOL. 1T218 I BLOW DOWN strainer as follows: (1) OPEN P44-F316 A(B,C), MECH VAC PMPA(B)(C) STR DR. (2) WHEN blowdown has been completed, THEN CLOSE P44-F316 A(B,C) MECH VAC PMPA(B)(C) STR DR. START MECH VAC PMP A(B)(C) with START pushbutton on 1H13 MCR 3 P680. CHECK proper vacuum pump operation for each running pump by 113' Turb A 3 Not OBSERVING the following: MVP Area Required Cooling Water Inlet Valve 1P44-SV-F514A, B, OR C has opened by 1T218 MOMENTARILY OPENING drain valve 1P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR. OBSERVING pressurized water flow, THEN CLOSE drain valve1 P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR. IF 1P44-SV-F514A, B, OR C did NOT open, THEN OPEN respective MECH VAC PMP A(B)(C) PSW SPLY BYP valve 1P44-F347A,B,C to provide cooling as needed for operation of Mechanical Vacuum Pump. I Suction Drain Valve SV-F507A, B OR Chas Closed by OBSERVING Page 257 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES no air suction flow.
Mechanical Vacuum Pump Inlet Valve F007A, B, OR Chas Opened. Proper oiler operation by OBSERVING oil flow from each oil return line. Secure Seal Steam Generator per 501 04-1-01-N33-1 PLACE Controller PK-R617 in MANUAL on 1H13-P878, AND CLOSE F506 as necessary to control reactor cooldown. The turbine Can be sealed with seal steam header pressure as low as 15 psig, PI-R622. Secure Reactor Feed Pump per 501 04-1-01-N21-1 IOI 03-1-01-4 Continued Off gas Preheater by placing controllers 1N64-R009A and 1N64-R009B Turbine 3 Not in manual and reducing output to O percent Building 93' Required Area 1 (1T113) Main Steam Isolation valves AND/OR Main Steam Line MCR 3 SHUTDOWN a Condensate s*ooster Pump AND CLOSE respective MCR 3 discharge valve per SOI 04-1-01-N19-1, leaving one Condensate Booster Pump in service SHUTDOWN a Condensate Pump AND CLOSE respective discharge MCR 3 valve per SOI 04-1-01-N19-1, leaving one Condensate Pump in service CLOSE B21-F069 MCR 3 OPEN the following MSIV drain valves:
- a. B21-F067A
- b. B21-F067B
- c. B21-F067C
- d. B21-F067D OPEN the following valves: MCR 3
- a. B21-F033
- b. B21-F068 ISOLATE extraction steam to the HP Feedwater heaters as follows: MCR 3 CLOSE the following valves:
- a. N36-F010A EXTR STM SPLY TO FW HTR SA
- b. N36-F01 OB EXTR STM SPLY TO FW HTR SB
- c. N36-F011A EXTR STM SPLY TO FW HTR 6A
- d. N36-F011 B EXTR STM SPLY TO FW HTR 68 Page 258 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES OBSERVE the following drain valves open: MCR 3 N36-F008A FW HTR 6A EXTR STM RO BYP DR VLV N36-F0088 FW HTR 68 EXTR STM RO BYP DR VLV 1 OPEN the following drain valves: MCR 3 OPEN HP Stop AND Control Valve Drain Valves by DEPRESSING each of the following MSCV UPSTRM DR VLV "JOG OPEN" pu.sh buttons:
- a. N33-F078A b.- N33-F0788
- c. N33-F078C')
- d. N33-F078D OPEN Left Side Crossover piping drains by DEPRESSING each of thei following XOVER PIPE LS DR VLV "JOG OPEN" pushbuttons:
- a. N11-F043A (FRIST)
- b. N11-F036A (FR 2ST)
- c. N11-F044A (RE IST)
- d. N11-F038A (RE2ST)
OPEN Right Side Crossover piping drains by DEPRESSING each of the following XOVER PIPE RS DR VLV "JOG OPEN" pushbuttons:
- a. N11-F0448 (FR IST)
- b. N11-F0388 (FR 2ST)
- c. N11-F0438 (RE IST)
- d. N11-F0368 (RE 2ST)
OPEN N11-F01S, MSCV A/8 DNSTRM DR VLV. OPEN the following MSR 2ND STG STM DRVLVS:
- a. N11-F301
- b. N11-F302 OPEN the following drain valves unless required closed to minimize MCR 3 cool down:
OPEN Main Stearn Line Drain Valves Nt1-FOS6, FOSS, F009, F011, F049, AND FOSO by DEPRESSING MSL DR LINE ISOL VLVS "OPEN" pµshbutton. OPEN MSL Bypass Drain valves (N11-F002A, F0028, F002C, F002D, F010, F007, FOS2A, FOS28, FOS7) using MSL DR VLVS DR LINE BYP. VLV "OPEN" pushbutton. DEPRESS Bqth NSSSS INBD ISOL RESET pushbutton (1H13-P601- MCR 3 188) AND NSSSSOTBD ISOL RESET pushbutton (1H13-P601-19B) to resetlogic AND re-energize RHR Logic lights on 1H13-P622 AND 1H13-P623 panels. TRANSFER to startup level control IF NOT already in service MCR 3 TRANSFER the RFPT A(B) SP CONT to MAN. MCR 3 Page 259 of 270
-===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES, IF MSIV's are open with Main Condenser available, THEN INITIATE MCR 3 AND MAINTAIN cooldown at s 90°F/hr with one of the following methods:
CONTROL Reactor cooldown with Manual Bypass Jack on 1H 13-P680-9C At approximately 200 psig Reactor pressure, SHUTDOWN one RWCU Pump per SOI 04-1-01-G33-1, IF Both are running. SLOWLY OPEN 1G33-F044, RWCU FLTR DMIN BYP VLV on 1H13- MCR 3 Not P680 while reducing F/D flow with flow controller 1G36-FC-R022A(B) Required CTMT 185' on 1G36-P002. RWCU Panel (1A509) MAINTAIN a nearly constant system flow rate, (450-500 gpm MCR 3 Is recommended), as indicated on 1G33-FI-R609, RWCU INL FLO, on1H13-P680. On 1G36-P002, OBSERVE that holding pump comes on WHEN F/D CTMT 185' 3 Not flow is< 80%. RWCU Panel Required (1A509) WHEN filter flow is< 20%, TURN Filter/Hold switch A(B) on 1G36-P002 CTMT 185' 3 Not to HOLD position. RWCU Panel Required OBSERVE the following valves fully Close: (1A509) G36-F001A(B) F/D Inlet G36-F002A(B) F/D Inlet G36-F003A(B) F/D Outlet G36-F004A(B) F/D Outlet OBSERVE HOLD light on AND FILTER light out on 1G36-P002 CTMT 185' 3 Not RWCU Panel Required (1A509) PLACE the MANUAUAUTO selector on controller 1G36-FC-R022A (B) CTMT 185' 3 Not in MANUAL position with controller output at 0% output. RWCU Panel Required (1A509) REPEAT Steps 4.6.2a AND 4.6.2b for second F/D. CTMT 185' 3 Not RWCU Panel Required (1A509) LOWER system flow rate to< 280 gpm by THROTTLING 1G33F044 as MCR 3 indicated on 1G33FI-R609, RWCU INL FLO, on 1H13-P680. TRIP one of the running RWCU pumps MCR 3 ESTABLISH 90 to 300 gpm flow as indicated on 1G33-FI-R609, MCR 3 RWCU INL FLO, on 1H13-P680 by THROTTLING the Bypass Valve 1G33F044 WHEN Reactor pressure is reduced to< 135 psig, THEN at Page 260 of 270
-===- Entergy .Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS approximately 40 psig, PLACE one loop of RHR System in SHUTDOWN COOLING mode per SOI 04-1-01-E12-2.
RA CK OUT RHR A/B PMP Breaker, 152-1509/1606 1 Control Bldg. 3 Required 111' SWGR Rms OC202, OC215 SHUTDOWN RHR JOCKEY PUMP A/B on 1H13-P871. MCR 3 CLOSE F082A/B, RHR JCKY PMP SUCT ISOL VLV, on 1H13-P871. MCR. 3 CLOSE F064A/B, RHR MIN FLO TO SUPP POOL. MCR 3 CLOSE F004A/B, RHR PMP SUCT FM SUPP \., MCR 3 ENSURE OPEN F003A/B, RHR HX OUTL VLV MCR 3 ENSURE OPEN F048A/B, RHR HX A BYP VLV. MCR 3 CLOSE F047A/B, RHR HX INL VLV. MCR 3 CLOSE F428A/B, PRESSURE LOCK ISOL for F024 RHRA/B . 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) CLOSE F438A/B, PRESSURE LOCK ISOL for F064 RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) SLOWLY OPEN F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC Rm (1A204) OPEN F006A, RHR PMP A SUCT FM SHUTDN CLG AND MONITOR MCR 3 RHR HR A STM press indicator for rise in pressure. VENT Shutdown Cooling suction header as follows: Aux Bldg 3 Required (a) OPEN F323. 119' RCIC Rm (1A204) (b) OPEN F399. (c) WHEN a solid stream of water is observed out ofvent line, THEN CLOSE F399. (d)
- CLOSE F323.
OPEN F073A, RHR HX A OTBD VENT VLV. MCR 3 OPEN F074A, RHR HX A INBD VENT VLV. MCR 3 VENT RHR A Heat Exchanger A as follows: Aux. 139' 3 Required (a) OPEN F400A, A RHR HX VENT. RHRA/B Rm Page 261 of 270
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES (b) OPEN F401A, A RHR HX VENT. 1A303, (c) WHEN water is observed from vent, THEN CLOSE F401A. 1A304/1A306 (d) CLOSE F400A. , 1A307 OPEN F064A/B. AFTER approximately one minute, THEN CLOSE MCR 3 F064A/B WHEN Conductivity as indicated on HX A/B OUT CNDCT, is as low as MCR 3 practical (Should be less than 2.0 µmhos/cm),
THEN CLOSE F073A/B, RHR HXA/B OTBDVENTVLV. CLOSE F074A/B, RHR HX A/B INBD VENT VLV MCR 3 LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC Rm (1A204) CLOSE F048A/B MCR 3 \ OPEN F063A/B, Manual Flush Valve. RHRA/B 3 Required Pump Rm Aux Bldg 119' (1A203/1A20 5) OPEN F073A/B, RHR HX A/B OTBD VENT VLV MCR 3 OPEN F074A/B, RHR HX A/B INBD VENT VLV MCR 3 WHEN Conductivity as indicated on HX A/B OUT CNDCT, is as low as MCR 3 practical (Should be less than 2.0 µmhos/cm), THEN CLOSE F073A/B, RHR HX A/B OTBD VENT VLV. CLOSE F074A/B, RHR HX A/B INBD VENT VLV MCR 3 LOCK CLOSED F063A/B, Manual Flush Valve. RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A203/1A20 5) OPEN F048A. MCR 3 OPEN F047A. MCR 3 ENSURE OPEN F003A. MCR 3 ENSURE Shutdown Cooling Isolation Logic is reset by PRESSING MCR 3 NSSSS INBD ISOL RESET pushbutton AND NSSSS OTBD ISOL RESET pushbutton on 1H13-P601. PLACE Standby Service Water A System in service to RHR A Heat MCR 3 Exchanger on 1H13-P870 as follows. START SSW Pump A per SOI 04-1-01-P41-1. Page 262 of 270
a-===-Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES OPEN P41-F014A, SSW INL TO RHR HX A. ENSURE OPEN P41-F068A, SSW OUTL FM RHR HX A. START RHR RM A FAN COIL UNIT. ENSURE OPEN F010, SHUTDN CLG MAN SUCT VLV. MCR 3 ENSURE CLOSED F040, RHR TO RADWST OTBD SHUTOFF VLV. MCR 3 ENSURE CLOSED F049, RHR TO RADWST INBD SHUTOFF VLV. MCR 3 OPEN F020, Manual Flush Valve approximately 3 turns. Valve May be Aux Bldg 3 Required opened further IF required for level control. 119' RCIC Rm (1A204) I OPEN F008, RHR SHUTDN CLG OTBD SUCT VLV MCR 3 OPEN F009, RHR SHUTDN CLG INBD SUCT \!LV as follows; MCR 3 ENSURE breaker 52-163137 is CLOSE position OPEN F009, RHR SHUTDN CLG INBD SUCT VLV MONITOR Reactor water level WHILE 1E12F009 AND 1E12F020 are OPEN. PERFORM IMMEDIATELY the next step 4.1.2.b(14) IF a rise in Reactor water level is NOT desired. J LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119' RCIC Rm (1A204) NOTIFY Radwaste Operators to be prepared for Reactor water flush to MCR 3 Required Waste Surge tank. Rad waste I Building 118' Radwaste Control Room (OR241) OPEN F203, RHR SYS FLUSH TO LIQ RADWST by the following MCR 3 handswitches to OPEN: F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-3C) F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C) ENSURE CLOSED F070A/B, Manual RHR Drain Valve Aux Bldg 93' 3 \_ Required Corridor (1A101) I OPEN F072A/B, RHR Drain Valve RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) Page 263 of 270 -, 1
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*===- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES SLOWLY OPEN F070A, RHR Drain Valve approximately one turn to Aux Bldg 93' 3 Required start flow to Radwaste. Corridor IF "RHR A DISCH PRESS ABNORMAL" annunciator alarms while (1A101) warming RHR A, THEN CLOSE F047A AND F048A to prevent draining of downstream piping.
THROTTLE F070A/B to warm RHR Pump A/Bat less than 100°F/hr Aux Bldg 93' 3 Required until RHR DISCH TO RADWST ON RHR TEMP recorder is 200°F OR Corridor within 100°F of RX water temp, whichever is less. (1A101) LOCK CLOSED F070A, RHR Drain Valve. Aux Bldg 93' 3 Required Corridor (1A 101) LOCK CLOSED F072A, RHR Drain Valve. RHRA/B 3 Required Pump Rm Aux Bldg 93' (1A 103/1A10 5) CLOSE F203, RHR SYS FLUSH TO LIQ RADWST by TAKING the MCR 3 following handswitches to CLOSE: F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-3C) F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C) RACK IN RHRA/B PMP Breaker, 152-1509/1606 Control Bldg. 3 Required 111' SWGR Rms OC202, OC215 NOTIFY Chemistry AND Radiation Protection that possibility of a MCR 3 crud burst Could occur due to starting of RHR pump in SOC mode START OR ENSURE running RHR RM A FAN COIL UNIT on 1H13- MCR 3 P870. ENSURE CLOSED F064A, RHR A MIN FLO TO SUPP POOL. MCR 3 ENSURE RHR JOCKEY PUMP A is shutdown. 'MCR 3 ENSURE CLOSED F082A, RHR A JCKY PMP SUCT ISOL VLV. MCR 3 ENSURE CLOSED F004A, RHR A SUCT FM SUPP POOL. MCR 3 ENSURE OPEN the following valves: MCR 3 (a) F010 (Concurrent Verification Required) (b) FOOS (c) F009 as follows; (1) ENSURE breaker 52-163137 is CLOSE position (2) ENSURE OPEN F009, RHR SHUTDN CLG SUCT VLV (d) F006A Page 264 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases
/
IOI / 501 ACTIONS LOCATION MODE NOTES (e) F047A (f) F048A CLOSE F003A, RHR HX A OUTL VLV. MCR 3 ENSURE CLOSED B21-F065A, FW INL SHUTOFF VLV. MCR 3 START RHR PMP A AND IMMEDIATELY FULLY OPEN one of MCR 3 the following valves: (a) E12-F053A, RHR A SHUTDN CLNG RTN TO FW (b) E12-F037A, RHR A TO CTMT POOL (c) E12-F042A, RHR A INJ SHUTOFF VLV MONITOR RHR HX A differential temperature on RHR MCR 3 TEMPERATURE RECORDER as follows: RHR HXA Point 1(inlet) - Point 5( outlet) ESTABLISH a cool down rate of less than 90°F/hr, as follows: MCR 3 Slowly JOG OPEN F003A to allow flow through heat exchanger, AND MONITOR cooldown rate. THROTILE one of the *following valves to maintain RHR pump flow -8600 gpm AND RHR heat exchanger flow -8200 gpm: IF flow is through F053A, THEN THROTILE F053A AS LONG AS flow through valve is maintained< 8550 gpm. IF E12-F003A is closed while in SHUTDOWN COOLING, MCR 3 THEN MONITOR REACTOR COOLANT TEMPERATURE using the following indications: REACTOR RECIRC LOOP A/8 suction temperature (IF recirc pump(s) running) RWCU REGENERATIVE HEAT EXCHANGER INLET temperature (IF RWCU pump(s) are running.) Point 5 of RHR TEMPERATURE RECORDER. Installed thermocouple suspended above Reactor core. WHEN F003A valve is t'ull open AND additional cooling is required, MCR 3 THEN SLOWLY THROTILE CLOSE F048A as needed to establish desired cooldown rate. WHEN F048A valve is full closed, THEN, IF desired, THROTILE MCR 3 F003A to MAINTAIN desired coolant temperature OR SOC flow while MAINTAINING ~ 3000 gpm flow. F048A may be fully opened to reduce ( cooldown rate but CANNOT be left in a throttled position UNTIL F003A is full open. SELECT "Shutdown Cooling-RHR A" OP GUIDE on PDS computer. MCR 3 The guide Should be left on-screen OR icon'd WHEN the respective shutdown cooling loop is in service until Reactor Coolant has been stabilized at desired temperature so that the guide Will warn operators IF Shutdown Cooling parameters are out of range Page 265 of 270
-::::=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES LOG Reactor coolant temperature on Data Sheet I of 03-1-01-3 OR MCR 3 other applicable IOI. TAKE temperatures as required by 03-1-01-3 during cooldown AND CONTINUE to take readings once per hour WHEN temperature is stable.
LOG temperatures for SSW/RHR HX AND reactor coolant on log MCR 3 similar to Attachment I to ENSURE SSW temperature does NOT exceed design temperatures. (Ref. CR 1997-0282) IF SSW A auto start signal from RHR A pump running is defeated by Temporary Alteration, THEN START/STOP SSW A AND B fans as necessary to MAINTAIN SSW A Supply temp. (E12-R601, pt. 12) between 50 AND 75 deg. IF RPV level control via RWCU blowdown is unavailable, MCR 3 THEN RPV level control May be established by USING E12-F073A AND E12-F074A RHR heat exchanger vent to establish RPV level control, AND THROTTLE OPEN E12-F073A AND E12-F074A as required to establish AND maintain the desired RPV level. MONITOR RPV level while reject is in progress. IF desired to add water to Reactor with SOC in operation WHEN in Aux Bldg 4, 5 Not Modes 4 OR 5, THEN PERFORM the following: 119' RCIC Required THROTTLE OPEN, F020. Rm (1A204) WHEN desired Reactor Vessel Level is reached, THEN LOCK CLOSED F020. IOI 03-1-01-3 Continued At approximately 120 psig, PERFORM the following: TRANSFER RWCU to Pre-pump mode per SOI 04-1-01-G33-1. PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601-19B to MCR 3 the TEST position. VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 the TEST position. VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13- P601-19A-G3) Alarms. SECURE RWCU blowdown flow per Section 5.1 of this instruction. MCR 3 STOP running RWCU pump AND leave F044, RWCU FLTR DMIN BYP MCR 3 VLV THROTTLED SLIGHTLY OPEN. '-. CLOSE the following valves AND proceed to Step 4.4.2g without delay: MCR 3 F250 RWCU SPLY TO RWCU HXS 1H13-P870-3C F251 RWCU SPLY TO RWCU HXS 1H13-P870-9C Page 266 of 270
Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 &H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES F252 RWCU HX RTN TO RWCU PMPS 1H13-P870-9C F253 RWCU HX RTN TO RWCU PMPS 1H13-P870-3C F255 RWCU FLTR/DMIN INL FM RWCU PMP 1H13-P870-5C OPEN OR CHECK OPEN the following valves: MCR 3 F004 PMP SUCT CTMT OTBD ISOL 1H13-P680 F001 PMP SUCT DRWL INBD ISOL 1H13-P680 F254 RWCU FLTR/DMIN INL FM RWCU HX 1H13-P870-5C F256 RWCU HX INL FM RWCU PMP 1H13-P870-5C CLOSE OR CHECK CLOSED F044, RWCU FLTR DMIN BYP VLV; MCR 3 And THEN RESTART one RWCU pump AND JOG OPEN F044 to establish flow greater than 90 gpm but less than 300 gpm. START one RWCU the pump AND THROTTLE F044 to achieve a MCR 3 system flow greater than 90 gpm, But less than 300 gpm as indicated on FI-R609, RWCU INL FLO. IF performing system warm-up. THEN MAINTAIN minimum flow, AVOIDING low flow trip. IF desired, START a second pump as follows: MCR 3 START the pump AND THROTTLE F044 to maintain 300 - 500 gpm system flow as indicated on FI-R609, RWCU INL FLO, with Both Pumps running. IF desired, ESTABLISH RWCU blowdown flow in accordance with All areas Section of this instruction previously addressed for this evolution IF desired, PLACE F/Ds in service in accordance with Section 4.5 of All areas this instruction. previously addressed for this evolution PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601-19B to MCR 3 the NORM position. VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Clears. PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601-18B to MCR 3 the NORM position. VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13- P601-19A-G3) Clears. IOI 03-1-01-3 Continued SHUTDOWN the running Condensate Booster Pump AND CLOSE respective discharge valve per SOI o4-1-01-N19-1. Page 267 of 270
-==~ Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS NOTES IF scheduled, THEN PERFORM 06-0P-1B21-R-0010 (Att. I AND/OR II) Not WHEN reactor pressure is between 50 AND 100 psig required to be performed to Shut down and Cool down the plant.
At approximately 60 psig Reactor pressure, PERFORM the following: MCR 3 VERIFY that RCIC system isolates automatically. IMMEDIATELY NOTIFY CAS, SAS, OR Security Island that RCIC is not available (non-functional). COMPLETE shutdown of RCIC system per SOI 04-1-01-E51-1. WHEN cooldown using Bypass Valves is no longer desired AND MCR 3 Shutdown Cooling is in service, THEN CLOSE the Bypass Valves as follows: SET the TURB STM PRESS DEMAND setpoint approximately 100 psig above Reactor pressure using the PRESS REF "RAISE" OR "LOWER" pushbuttons on 1H13-P680-9C. DEENERGIZE the Manual Bypass Valve Controller by depressing the MAN BYP CONT "OFF" pushbutton on 1H13-P680-9C. IF MSIV's are open AND stroke time testing was NOT scheduled, MCR 3 THEN PERFORM the following: CLOSE the following Inboard MSIVs: B21-F022A B21-F022B B21-F022C B21-F022D WHEN Main Steam Line pressure downstream of MSIVs is near zero psig, THEN CLOSE the following Outboard MSIVs: B21-F028A B21-F028B B21-F028C B21-F028D CLOSE B21-F016 CLOSE B21-F019 NOTIFY Radiation Protection that the Reactor is to be vented to MCR 3 Drywell sump AND REQUEST Drywell survey after Head Vent realignment. WHEN Reactor coolant temperature is less than 210°F, THEN MCR 3 REALIGN Reactor Head Vents on 1H13-P601 as follows: OPEN 1B21-F001, RPV OTBD VENTVLV. Page 268 of 270
~Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES PEN 1 B21-F002, RPV INBD VENT VLV.
CLOSE 1821-FOOS, RPV VENT TO MSL A. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is n1ot included in this assessment or in Table H-2. Page 269 of 270
~=- Entergy Grand Gulf Nuclear Station EAL Basis Document Revision XXX Attachment 2 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Table A-3 & H-2 Results Table A-3 & H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Control Building 111' SWGR Rnis (OC202, OC215) 3 Auxiliary Building 93' RHR A Pump Room (1A 103) 3 Auxiliary Building 93' RHR B Pump Room (1A105) 3 Auxiliary Building 93' Corridor (1A101) 3 Auxiliary Building 119' Corridor (1A201) 3 Auxiliary Building 119' RHR A Pump Room (1A203) 3 Auxiliary Building 119' RHR B Pump Room (1A205) 3 Auxiliary Building 119' RCIC Room (1A204) 3 Auxiliary Building 139' RHR A Room (1A303, 1A304) 3 Auxiliary Building 139' RHR .B Room (1A306, 1A307) 3 Radwaste Building 118' Radwaste Control Room (OR241) 3 Page 270 of 270}}