GNRO-2004/00057, Response to Request for Additional Information for Proposed Upgraded Emergency Action Levels (Eals) Using NEI 99-01 Revision 4 Methodology
| ML043280559 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/19/2004 |
| From: | Bottemiller C Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GNRO-2004/00057 | |
| Download: ML043280559 (287) | |
Text
GNRO-2004/00057 November 19, 2004 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
Grand Gulf Nuclear Station - Response to Request for Additional Information for Proposed Upgraded Emergency Action Levels (EALs)
Using NEI 99-01 Revision 4 Methodology Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 REFERENCES : 1 . 6/23/2004 e-mail from T. Alexion, NRC, to L England, EAL RAls for GG and RB
- 2. December 16, 2003 letter to Document Control Desk, Grand Gulf Nuclear Station - Proposed Upgraded Emergency Action Levels (EALs) Using NEI 99-01 Revision 4 Methodology - GNRO-2003/00067
- 3. NEI 99-01, Rev 4, "Methodology for Development of Emergency Action Levels".
or Madam:
Reference 2 provided the Grand Gulf Nuclear Station (GGNS), Unit I submittal of proposed EALs using the methodology outlined in NEI 99-01, "Methodology for Development of Emergency Action Levels" (Revision 4, January 2003). Reference 1 contained the NRC Requests Additional Information. This letter provides GGNS response to Reference 1 . In responding the RAls Entergy took the opportunity to further standardize our regional approach and made format and editorial changes in addition to addressing the RAI issues .
Accordingly, a complete revision to our initial submittal is enclosed that incorporates all changes as described in the attachments to this letter .
Significant changes have been reviewed and approved by the Onsite Safety Review Committee. Prior to implementation changes will be discussed and agreed upon with state authorities as required .
GNRO-2004/00057 Page 2 Plant specific information is attached as follows:
Response to NRC RAI Questions GGNS Cross Reference Matrix from NEI EAL number to Entergy EAL number (i .e.
NEI number, previous Entergy number, new Entergy number)
Proposed EALs - To Be Incorporated in Procedure Proposed EAL Bases - To Be Incorporated in Procedure NEI 9901, Rev. 4 to Plant Specific Correlations, Differences, Deviations, and Justifications Differences and Deviations from NEI 99-01 Rev 4 are based on NRC guidance contained in Supplement I to RIS 2003-00018 dated July 13, 2004.
We request NRC approval of this submittal within 120 days of receipt. GGNS plans to implement these new EALs at the earliest opportunity following NRC approval.
If you have any questions regarding this submittal, phase contact Mr. Matt Ckawford at 601-437-2334 Sincerely, CAB/MLC:mlc Attachments:
1 . Response to NRC RAI Questions
- 2. GGNS Cross Reference Matrix
- 3. Proposed EALs - To Be Incorporated in Procedure
- 4. Proposed EAL Bases - To Be Incorporated in Procedure
- 5. NEI 99-01, Rev. 4 to Plant Specific Correlations, Differences, Deviations, and Justifications Enclosures :
1 . State and Local Concurrence
- 2. Simplified Drawing of Distribution Switchyard
- 3. Basis for Radiological Effluent Initiating Conditions
- 4. Safety Analysis Parameters in the Emergency Action Limits cc: (See Next Page)
GNRO-2004/00057 Page 3 cc: w/a (except as noted)
NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S. Nuclear Regulatory Commission ATTN: Dr. Bruce S. Mallett (w/2)
Regional Administrator, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U.S. Nuclear Regulatory Commission ATTN: Mr. Nageswaran Kalyanam, NRR/DLPM (w/2)
ATTN: ADDRESSEE ONLY ATTN: U.S. Postal Delivery Address Only Mail Stop OWFN/7D-1 Washington, D.C. 20555-0001 Mr. D. E. Levanway (Wise Carter)
Mr. L. J. Smith (Wise Carter)
Mr. N. S. Reynolds Mr. H. L. Thomas
Attachment 1 To GNRO-2004/00057 Response to NRC RAI Questions to GNRO-2004/00057 Page 1 of 17 REQUESTS FOR ADDITIONAL INFORMATION (RAIs)
REGARDING ADOPTION OF NEI 99-01, REVISION 4 FOR GRAND GULF NUCLEAR STATION, UNITS 1 DOCKET NUMBER 50-416 By letter GNRO-2003/00067, dated December 16, 2003 (ML033510722), Entergy Operations, Inc. submitted proposed changes to the emergency action levels (EALs) for Grand Gulf Nuclear Station (GGNS), Unit 1. This submittal revises the GGNS EALs from the current NUREG-0654, Appendix 1 scheme to NEI 99-01, Revision 4.
The NRC staff performed a review on the proposed EAL Matrix (Attachment 3), EAL Basis (Attachment 4), and Plant-Specific Correlations, Differences, Deviations and Justification (Attachment 5). Proposed changes to the GGNS FSAR, contained in Attachments 1 and 2, were not included as part of the NRC staffs review.
The NRC staff has the following questions related to this submittal:
General Comments:
- 1. 10 CFR 50, Appendix E -- Section IV.B (Assessment Actions) states, ...emergency action levels shall be discussed and agreed on by the applicant [licensee] and State and local governmental authorities, and approved by NRC. (Italics added) Submittal cover letter reflects that these changes have been reviewed and approved by the States of Mississippi and Louisiana, and local government authorities.
Provide documentation of when these discussions occurred and agreement by the State and local governmental authorities on the implementation of the proposed EAL changes based on NEI 99-01, Revision 4. Specific reference needs to be made regarding State and local concurrence for any proposed changes that would require either a higher or lower classification than that designated in the NEI 99-01 guidance.
GGNS Response Copies of documents indicating review and concurrence with the originally proposed EAL changes based on NEI 99-01, Revision 4 by the following state and local government authorities are provided as Enclosure 1.
- Mississippi Emergency Management Agency
- Mississippi State Department of Health (MSDH)
- Claiborne County Civil Defense
- Tenesas Parish Office of Homeland Security & Emergency Preparedness
- Louisiana Office of Homeland Security and Emergency Preparedness
- 2. Provide update to Attachment 5 (Plant Specific Correlations, Differences, Deviations and Justifications),
based on an evaluation of changes proposed to NEI 99-01 guidance, to ensure that any deletions to NEI 99-01 Initiating Condition (IC) statements, example EALs criterion and basis, or significant content changes (other than format, nomenclature, simple terminology or system names, etc.) that may impact intent or thresholds established or guidance provided in NEI 99-01, are listed as deviations. In addition, provide site-specific technical justification for any deviations, as appropriate. (Specific examples are listed under Specific Comments, but are not all inclusive.)
GGNS Response The Deviations and Differences document (Attachment 5) was updated.
to GNRO-2004/00057 Page 2 of 17
- 3. Provide a simplified drawing or schematic illustrating unit auxiliary and start-up transformers, and describe the inter-relationship regarding conditions needed for a loss of off-site power and the ability of emergency diesel generators to supply on essential busses.
GGNS Response A simplified electrical distribution drawing is provided as Enclosure 2, illustrating unit auxiliary and start-up transformers, and showing the inter-relationship regarding conditions needed for a loss of off-site power and the ability of emergency diesel generators to supply on essential busses. The drawing is for information only and is not intended for official use.
- 4. Provide a copy, or include a detailed description in Attachment 4, of calculations used to determine effluent monitor thresholds under AG1, AS1, AA1 and AU1, and specify any deviations in Attachment 4 from guidance contained in Appendix A to NEI 99-01 (Basis for Radiological Effluent Initiating Conditions).
GGNS Response A description of the method use to determine setpoints is included as Enclosure 3. Deviations are specified in Attachment 5.
to GNRO-2004/00057 Page 3 of 17 Specific Comments:
- 1. AU1 (corresponds to NEI 99-01, AU1)
AA1 (corresponds to NEI 99-01, AA1)
- a. Licensee applies EAL 1 only to liquid releases. Identify as a deviation in Attachment 5 and provide technical justification, or provide proposed change to comply with NEI 99-01, AU1 (AA1)
/ EAL 1 wording.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01. GGNS is only allowed to discharge liquids in accordance with the State Discharge Permit and per GGNS procedures.
- b. Licensee applies EAL 2 only to gaseous releases and does not provide monitor thresholds for an inadvertent liquid releases that might occur outside of a planned discharge. Identify as a deviation in Attachment 5 and provide technical justification, or provide proposed change to comply with NEI 99-01, AU1 (AA1) / EAL 2 wording.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01. GGNS has no effluent monitors on non-routine pathways. Any other unplanned release to the environment, such as a spill of contaminated liquid, would be identified through the use of sample analysis via EAL #3.
- 2. AA1 (corresponds to NEI 99-01, AA1)
The deviation using a threshold of 20 times ODCM limit appears justified; however, as part of response to General Comment #1, describe methodology used to ensure consistency with NEI 99-01, AU1/AA1 -
>AS1/AG1 and Appendix A guidance. Specifically evaluate calculations against those proposed by River Bend Station (RBS), which uses 200 times TS / ODCM limit as threshold.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01.
- 3. AA2 (corresponds to NEI 99-01, AA3)
- a. Describe the methodology used for determining areas requiring continuous occupancy to maintain plant safety functions in EAL 1. Specifically, provide justification for not including other control stations that may be manned continuously (i.e., per NEI 99-01 AA3 Basis, the radwaste control room and central alarm station). In addition, describe how these areas requiring continuous occupancy correlate to those identified by other Entergy stations under their respective NEI 99-01 submittals.
GGNS Response The location of control stations is unique to each Entergy site. The central alarm station is not listed since it is located within the control room envelope. The radwaste control room does not have to be continuously manned.
- b. Provide technical rational in licensee Basis for selection of > 80 R/hr threshold in EAL 2.
Describe why threshold would not impede operator access based on NEI 99-01 Basis guidance, and how threshold correlates to station normal occupational guidelines and limits, versus emergency access under EOPs. In addition, describe how this threshold correlates to those identified by other Entergy stations under their respective NEI 99-01 submittals.
to GNRO-2004/00057 Page 4 of 17 GGNS Response The initiating condition for EAL 2 is impedes operation of systems required to maintain safe operations or is to establish or maintain cold shutdown. The NEI bases includes impede, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant. The value was revised to 10 R/hr in Attachments 3, 4, and 5 to be consistent with Entergy Nuclear South (ENS) sites submitting EAL revisions for approval.
A standard set of radiological practices and procedures does exist for ENS plants contained in corporate Radiation Protection Procedures. These procedures do require specific actions prior to an expected dose of 5 Rem. With regard to application to this EAL, these procedures are in general based on expected dose for an activity and not exposure rates. The use of 10 R/hr would mean a worker could perform unimpeded activities in the area for 30 minutes. RP-103, Access Control, step 5.1.4 states Stay times are required for activities that will result in an exposure of > 500 mrem/entry or will be performed in posted LHRA or VHRAs. NEI 99-01 states As used here, impede, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant. This guidance implies that some actions required by exposure rates encountered may not be severe enough to warrant considerations as applicable to this IC because they may not represent a significant (emphasis added) threat to the safe operation of the plant. Exposure rates and required time in the area that together cause the requirement to use stay times do not appear in themselves enough to meet the NEI criteria as stated above. Neither does any requirement for briefings as these would be expected to occur regardless in order to conduct the activities required with the Radiation Protection briefing included as a part of the task briefing. Therefore, Entergy establishes a value for this EAL that considers stay times that may require entries with multiple personnel to accomplish a task to prevent exceeding Entergy administrative limits or will require extension of the administrative limits.
- c. Licensee EAL 2 criteria states that access is required. Per NEI 99-01 guidance, access does not need to be required, but rather access to areas requiring infrequent access to maintain plant safety functions would be impeded. Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01.
- 4. AA3 (corresponds to NEI 99-01, AA2)
...outside the reactor vessel. Provide proposed change to address inconsistency in terminology.
GGNS Response The IC statement in Attachments 3 and 5 was revised to state, outside the reactor vessel.
- b. Under EAL 1, licensee included qualifier, potential spent fuel damaging event. Describe the criteria that will be used by licensee, in addition to a valid radiation monitor alarm, to determine when qualifier is met. In addition, describe how this threshold correlates to those identified by other Entergy stations under their respective NEI 99-01 submittals.
to GNRO-2004/00057 Page 5 of 17 GGNS Response The qualifier potential spent fuel damaging event has been deleted to be consistent with NEI 99-01.
- c. Describe how radiation monitors proposed in licensee EAL 1 correlate to radiation monitor locations listed in NEI 99-01 AA2 / EAL 1 criterion.
GGNS Response The GGNS radiation monitors cover similar areas to those listed in NEI 99-01 AA2, EAL 1.
- d. Identify revisions to licensee EAL 2 as a deviation to NEI 99-01 AA2 / EAL 2 criterion in Attachment 5. Clarify whether a method may be available during refueling outages where level can be monitored or alarm capability provided.
GGNS Response Changes to EAL 2 are identified as a deviation in Attachment 5. The most valid method to monitor level during plant operations or during a refueling outage is via personnel observation.
In addition, alarm capability would be provided from area radiation monitors.
- a. Use of qualifier, using actual meteorology, in IC statement is inappropriate based on calculation of effluent monitor thresholds per NEI 99-01 AS1(AG1) / EAL 1, which requires the use of annual average meteorology. Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The qualifier using actual meteorology was removed from AS1 to comply with NEI 99-01 guidance. The qualifier remains with AG1 as specified in NEI 99-01.
- b. Monitor readings are intended to be calculated based on exceeding IC limit for each release monitor. Assessment of release from multiple release points would be provided via dose projection. Provide radiation monitor thresholds for effluent release points reflecting calculation methodology described in NEI 99-01 AS1 and Appendix A for exceed IC limit on each respective release point. In addition, describe how these thresholds correlate to those identified by other Entergy stations under their respective NEI 99-01 submittals.
GGNS Response Radiation monitor thresholds for effluent release points are provided in Table R1 in Attachments 3 & 4. These thresholds were calculated in accordance with the methodology provided in response to general question 4. These thresholds correlate directly with the other Entergy submittals.
- c. Provide proposed change for licensee EAL 1 to address NEI 99-01, AS1(AG1) / EAL 4 criteria for field survey results (i.e., closed window dose rates exceeding 100 mR/hr (AS1) / 1000 mR/hr (AG1) expected to continue for more than one hour, OR sample analysis indicate thyroid CDE of 500 mR/hr (AS1) / 5000 mR/hr (AG1) for one hour one inhalation).
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01.
to GNRO-2004/00057 Page 6 of 17
Provide statement under justification in Attachment 5 that supports use of units in uCi/ml as equivalent to NEI 99-01 threshold in Ci/gm.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01.
Licensee threshold for a LOSS and Potential LOSS incorrectly states that reactor water level cannot be restored above specific threshold. Per NEI 99-01, exceeding threshold alone is sufficient to consider either a loss or potential loss of fuel clad barrier. Provide proposed changes to comply with NEI 99-01 guidance of Level LESS THAN (site-specific) value.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI 99-01.
- a. NEI 99-01 Basis guidance states that reading should indicate 300 uCi/gm I-131 or the calculated concentration equivalent to the clad damage used in FC1. NEI 99-01, Fuel Clad Barrier Example EAL #1, further states that value corresponds to less than 5% fuel clad damage; however, licensee FC3 Basis in Attachment 4 states approximately 5%. Provide proposed change to Attachment 4 to reflect NEI guidance on basis for EAL threshold. In addition, discuss rational for not including CAUTION statement from NEI 99-01 regarding effects of shine.
GGNS Response The EAL was revised in Attachments 3 & 4 to be consistent with NEI 99-01.
- b. Provide reference in Attachment 4 to site-specific calculations performed to determine 5,000 R/hr threshold. In addition, provide a copy of calculations for review.
GGNS Response The threshold was changed to 3000 R/hr and explained in the bases (Attachment 4). A copy of the appropriate calculation is provided as Enclosure 4.
Provide technical justification that Main Steam Line Radiation levels > Hi-Hi Alarm Set Point is equivalent to 300 uCi/gm dose equivalent I-131, which serves as the basis for fuel clad barrier LOSS due to primary coolant activity (Example EAL #1) and drywell radiation monitoring (Example EAL #3).
GGNS Response The EAL was deleted to be consistent with NEI 99-01.
Addition of qualifier, ...with indications of a leak in the drywell, is not provided as a deviation. Identify as a deviation in Attachment 5.
GGNS Response The qualifier ...with indications of a reactor coolant leak in the drywell, was added as a deviation in Attachment 5.
to GNRO-2004/00057 Page 7 of 17
- a. Licensee threshold for a LOSS incorrectly states that reactor water level cannot be restored above specific threshold. Per NEI 99-01, exceeding threshold alone is sufficient to consider a loss of the fuel clad barrier. Provide proposed changes to comply with NEI 99-01 guidance of Level LESS THAN (site-specific) value.
GGNS Response The statement was deleted and Attachments 3, 4, and 5 were revised to comply with NEI.
- b. Licensee states in Attachment 5 that qualifier, with indications of a leak in the drywell, was added to ensure consistency with RCS Barrier EAL 1 (Drywell Pressure). This application is incorrect, since LOSS criterion is intended to stand alone for classifications purposes. RPV Level reaching TOAF is intended to cover scenarios that result not only due to an inventory loss, but as a result of a loss of makeup that would lead to core uncovery. Provide proposed change to remove qualifier, with indications of a leak in the drywell.
GGNS Response The qualifier with indication of a reactor coolant leak in the drywell was added as a deviation in Attachment 5.
- c. In Attachment 5, under differences, the licensee incorrectly lists -192 in. (MSCRWL) as threshold for barrier LOSS. This is inconsistent with threshold of -162 in. (TOAF) used in Attachments 3 and 4, and NEI 99-01 guidance. Provide proposed change to address inconsistency and comply with NEI 99-01 guidance.
GGNS Response The inconsistency has been corrected [i.e., -192 in. (MSCRWL) was removed].
- a. NEI 99-01 Basis for RCS Leak Rate specifically states [a]n unisolable main steam line (MSL) break. Identify as a deviation in Attachment 5 and provide justification for including an RCIC steam line break as a RCS barrier LOSS. In addition, provide site-specific indication of an isolable main steam line break to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to be consistent with NEI for LOSS.
- b. Provide basis for listing MSL Pipe Tunnel Temp and MSL Rad Monitor as criteria for a Potential LOSS when this appears to be in conflict with proposed licensee LOSS criteria due to a MSL or RCIC steam line break.
GGNS Response MSL Pipe Tunnel Temp and MSL Rad Monitor as criteria for a Potential LOSS were removed.
- c. Identify as deviation and provide technical justification for removing NEI 99-01 qualifier inside drywell under 1st Potential Release for RCS leakage greater than 50 gpm.
GGNS Response The qualifier inside drywell added to first potential loss criteria for RC3.
to GNRO-2004/00057 Page 8 of 17
- d. Under 2nd Potential Loss criteria, licensee inserts qualifier with confirmed Reactor Pressure Boundary leakage in the area, rather than NEI 99-01 criteria of unisolable primary leakage outside drywell. Licensee change is inconsistent with NEI 99-01 guidance where area and temperature levels are used to confirm a leak outside drywell (containment). Provide proposed change to comply with NEI 99-01, RCS Barrier Example EAL #3 criterion.
GGNS Response Second potential loss criteria changed to be consistent with NEI 99-01.
- e. In Attachment 5, under EAL #3, the licensee does not address 2nd Potential LOSS criteria as listed in Attachments 3 and 4. Provide proposed change to address inconsistency.
GGNS Response The EAL was revised in Attachment 5 to comply with NEI 99-01.
Provide further technical justification for elimination of drywell (containment) radiation monitor threshold based on applicability of other BWR-6 stations, which have included drywell (containment) radiation monitor in NESP-007 or proposed NEI 99-01 EALs. Clarify why threshold cannot be calculated based on RCS leakage with normal RCS coolant activity (i.e., value calculated would be less than radiation monitor reading recorded during normal plant operation).
GGNS Response Grand Gulf added an EAL to comply with NEI 99-01. The threshold for this EAL is discussed in Attachment 4.
Describe in Attachment 4 the justification for why a stuck open relief valve alone, without a corresponding increase in Suppression Pool bulk temperature approaching allowable Technical Specification limits or abnormal coolant activity, provides sufficient technical justification to warrant an Alert declaration.
GGNS Response The stuck open relief valve was deleted from the EALs to be consistent with NEI 99-01.
- a. Licensee does not address NEI 99-01 LOSS criteria, Drywell pressure response not consistent with LOCA conditions, nor identifies as a deviation in Attachment 5. This criteria is intended to address a condition were a loss of drywell (containment) exists at the time of a LOCA where pressure does not increase as expected. Provide proposed change to comply with NEI 99-01 guidance based on pressure in containment per Mark III design.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- b. Use of 22 psig Containment pressure is inconsistent with NEI 99-01 criterion submitted by Industry and approved by NRC under associated safety evaluation. Provide proposed change to comply with NEI 99-01 criterion reflecting a potential loss of primary containment based on design pressure (15 psig).
GGNS Response to GNRO-2004/00057 Page 9 of 17 The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
Per NEI 99-01 guidance, a Potential LOSS of primary containment is appropriate when an explosive mixture exists (above deflagation limits), regardless of hydrogen ignitor status. Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- a. NEI 99-01 criterion was not intended to consider the Primary Containment Barrier lost based on a failure to isolate alone. Identify as deviation in Attachment 5, why licensee did not include NEI 99-01 qualifier, AND downstream pathway to the environment exists, under 1st LOSS criterion and provide justification for consideration.
GGNS Response The EAL was revised to comply with NEI 99-01 guidance.
- b. Under 2nd Loss criterion, licensee inserts qualifier with confirmed Reactor Pressure Boundary leakage in the area, rather than NEI 99-01 criteria of unisolable primary leakage outside drywell. Licensee change is inconsistent with NEI 99-01 guidance where area and temperature levels are used to confirm a leak outside drywell (containment). Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised to comply with NEI 99-01 guidance.
- c. In Attachment 5, under EAL #4 description, the licensee does not list 3rd LOSS criterion (unisolable primary leakage outside drywell/containment), which is contained in Attachments 3 and 4. Provide proposed change to address inconsistency.
GGNS Response The inconsistency regarding the third loss criterion was addressed in Attachments 3, 4, and 5 consistent with the NEI 99-01 guidance.
- 18. PC5 (corresponds to NEI 99-01, Table 5-F-2: Containment Barrier Example EAL #4) Provide rational in Attachment 4 consistent with NEI 99-01 guidance for drywell radiation monitor reading (i.e., clad damage of 20%). In addition, provide a copy of calculations for review.
GGNS Response The Containment radiation monitor reading was changed. The rationale for the reading chosen is provided in Attachment 4. A copy of the associated calculation is provided as enclosure 4.
- 19. PC?? (corresponds to NEI 99-01, Table 5-F-2: Containment Barrier Example EAL #5) Provide discussion in Attachment 5 of evaluation performed to identify other site-specific indications of a loss or potential loss of the Containment Barrier per NEI 99-01 guidance.
GGNS Response The results of an evaluation performed are included in Attachment 5.
- 20. HU1 / EAL 2 (corresponds to NEI 99-01, HU4 / EAL 2) to GNRO-2004/00057 Page 10 of 17 Licensee EAL 2 defines a Site Security Code YELLOW as an armed adversary attempting to or has entered company property. NEI 99-01 guidance states, Security events as determined from site-specific Safeguards Contingency Plans, which per the Attachment 4 Basis discussion includes sabotage, hostage / extortion, civil disturbance, and strike action. Provide proposed change to EALs to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised to be consistent with the NEI 99-01 guidance.
HA4 / EAL 2 (corresponds to NEI 99-01, HA1 / EAL 2)
Provide further justification why wind speed defined under FSAR design tornado should not be used to meet NEI 99-01 criteria for high winds. Describe at what recorded wind speed would a tornado exist.
GGNS Response The NEI guidance relates to damage resulting from tornadoes or high winds. To address the question, Grand Gulf divided the EAL describing the two phenomena into separate EALs. One addresses a tornado striking within the Protected Area. The other covers severe weather with indication of sustained winds greater than or equal to 74 miles per hour within the Protected Area. Together the two EALs meet the intent of the NEI guidance and are further defined in Attachment 4.
- 22. HU3 (corresponds to NEI 99-01, HU1)
- a. Provide further justification why wind speed, defined under FSAR design tornado, should not be used to meet NEI 99-01 guidance. If change retained, identify deletion of high winds from licensee EAL 2 criteria as a deviation in Attachment 5.
GGNS Response Grand Gulf divided EAL 2 into separate EALs addressing tornadoes and high winds. The changes meet the intent of the NEI 99-01 guidance. Also, refer to the response to question 21 above.
- b. Provide proposed change to licensee EAL 4 to define damage in accordance with NEI 99-01, Section 5.4 definition of visible damage.
GGNS Response The definition section of the EAL implementing procedure will contain NEI 99-01 definitions when EALs are approved and procedure is issued.
- c. Describe evaluation performed to identify site-specific phenomena (i.e., low/high river level or other FSAR external hazards, etc.). Provide proposed change to address site-specific phenomena, or identify and justify as deviation in Attachment 5.
GGNS Response EAL 6 was added to address flooding.
to GNRO-2004/00057 Page 11 of 17
HA4 (corresponds to NEI 99-01, HA1 / EAL 5)
Provide indication in licensee HU3 for uncontrolled internal flooding in areas where flooding could potentially affect safety-related equipment as defined under EOPs per NEI 99-01 guidance. This EAL would subsequently escalate to an Alert under NEI 99-01, HA1 / EAL 5 if water level reached Maximum Safe Operating Value or equivalent. In addition, address under HA4 internal flooding events that create industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.
GGNS Response An EAL has been added to address internal flooding consistent with the NEI 99-01 guidance.
- 24. HU4 (corresponds to NEI 99-01, HU2)
As proposed, licensee EAL appears to be limited to fires within buildings containing plant vital areas.
Licensee incorrectly limits fires to within the power block structure. Provide proposed change to address NEI 99-01 criteria for a fire within the protected area that also applies building and areas contiguous (in actual contact or immediately adjacent) to plant vital areas.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- 25. HU5 (corresponds to NEI 99-01, HU3)
- a. Provide definition in licensee Basis, or other applicable plan section or procedure, that defines normal plant operation consistent with that defined in Section 5.4 to NEI 99-01.
GGNS Response The definition section of the EAL implementing procedure will contain NEI 99-01 definitions when EALs are approved and procedure is issued.
- b. Licensee justification for deletion of NEI 99-01, HU3 / EAL 2 is that no industries are near the GGNS site. This rational does not address possible transportation accidents in the vicinity of the GGNS site. Provide proposed change to comply with NEI 99-01, HU3 / EAL 2 criterion that event classification is warranted based on a report by respective offsite agencies of the need for the evacuation or sheltering of site personnel based on an offsite event.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
NEI 99-01 guidance states, Other security events as determined from site-specific Safeguards Contingency Plans, which per Attachment 4 Basis guidance includes sabotage, hostage / extortion, and strike action. Licensee EAL criteria is limit to addressing an intrusion by a hostile force into the protected area. Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- a. Provide proposed change to licensee EAL 2 to address visible damage to plant structures (containing systems and functions required for safe shutdown), in lieu of damage to a vital area that is contained within structure. In addition, address NEI 99-01, HA1 / EAL 2 and 3 criterion for events resulting in indication of degraded performance.
to GNRO-2004/00057 Page 12 of 17 GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- b. Define what plant structures are within the vital area.
GGNS Response The affected plant structures are listed in Table H2 of Attachments 3 & 4.
HA5 (corresponds to NEI 99-01, HA2)
Provide proposed change to address NEI 99-01, HA2 criterion, AND Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equipment within specified areas (required to establish or maintain safe shutdown).
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
In Attachments 3 and 5, licensee EAL 5 states, ...resulting in visible damage to safety related equipment. However, licensee EAL 5 description in Attachment 5 states, ...resulting in visible damage to vital area. Provide proposed change to address inconsistency per NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- a. Licensee revised IC statement limiting classification to events affecting the ...operation of safety systems required to establish or maintain safe shutdown, which fails to address requirement to maintain safe operation. Provide proposed change to IC statement to address NEI 99-01 criteria covering the operation of systems required to maintain safe operation or establish and maintain safe shutdown.
GGNS Response The IC was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- b. Licensee EAL is limited to gases within a vital area, which is more restrictive than NEI 99-01 criteria. Provide proposed change to address NEI 99-01 criterion covering the report or detection of gases within or contiguous to a vital area.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
NEI 99-01 guidance states, Other security events as determined from site-specific Safeguards Contingency Plans, which per Attachment 4 Basis discussion includes sabotage and hostage /
extortion. Licensee EAL criteria only addresses an intrusion by a hostile force into the plant vital area.
Provide proposed change to comply with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
to GNRO-2004/00057 Page 13 of 17
- 32. HG1 (corresponds to NEI 99-01, HG1)
- a. Provide proposed change to comply with NEI 99-01 Basis guidance that safety functions for a BWR include decay heat removal (ability to maintain a heat sink).
GGNS Response The EAL was revised in Attachment 4 to address decay heat removal (ability to maintain a heat sink) in compliance with NEI 99-01.
- b. Provide proposed change to Attachment 4 Basis discussion to include NEI 99-01 guidance, which states, This EAL should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly off-loaded reactor core in pool).
GGNS Response The EAL was revised in Attachment 4 to address loss of physical control of spent fuel pool cooling systems in compliance with NEI 99-01.
- 33. SU1 (corresponds to NEI 99-01, SU1 and CU3)
- a. Identify as a deviation in Attachment 5 and provide technical justification for deleting NEI 99-01 criterion, AND At least (site-specific) emergency generators are supplying power to emergency busses.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- b. Correct incomplete sentence under Differences discussion in Attachment 5.
GGNS Response The topic of the incomplete sentence is no longer in the Differences discussion.
- 34. SU3 (corresponds to NEI 99-01, SU5)
Identify the revision to threshold for identified leakage as a deviation and provide justification in Attachment 5.
GGNS Response The revision to threshold for identified leakage is described and justified as a deviation in Attachment 5.
- 35. SU4 (corresponds to NEI 99-01, CU1)
Provide further technical justification for elimination of NEI 99-01 criteria, based on other capabilities to monitor RCS leakage will in Modes 4 and 5 (i.e., sump in-leakage, etc.). Specifically justify why in-leakage cannot be monitored, but is utilized in NESP-007 and proposed NEI 99-01 schemes at other Industry BWR-6 stations.
GGNS Response An explanation why in-leakage cannot be monitored is provided in Attachment 5.
Identify deletion of NEI 99-01 EAL 2 (positive startup rate) as a deviation, rather than difference, in Attachment 5.
GGNS Response Deletion of NEI 99-01 EAL 2 (positive startup rate) is identified as a deviation in Attachment 5.
to GNRO-2004/00057 Page 14 of 17
- a. Licensee includes cellular (cell) telephones under offsite communications capability. Clarify in Attachment 5 whether implementing procedures address the use of cellular phones as a back-up means of offsite communications. In addition, confirm that cellular phones will function within or in close proximity to plant structures to be considered an effective means of offsite communications. Provide a proposed change to delete cellular phones if its use is not proceduralized, since this should be considered an extraordinary means.
GGNS Response Cellular phones were deleted from the list since it is not a backup in accordance with approved procedures.
- b. Licensee lists the UHF radios under offsite communications equipment, but NEI 99-01 Basis describes radio transmissions as an extraordinary means of offsite communications. Clarify in Attachment 5 whether implementing procedures address the use of the Radio System as a back-up means of offsite communications. Provide a proposed change to delete UHF radios if its use is not proceduralized, since this should be considered an extraordinary means.
GGNS Response One method of backup communications listed in the procedures is the UHF radio. The Emergency Plan Procedure, 10-S-01-6, Notification of Offsite Agencies and Plant On-Call Emergency Personnel, addresses the use of the UHF radio as backup communications.
- c. Licensee lists satellite phones under offsite communications equipment. Clarify in Attachment 5 whether implementing procedures address the use of the satellite phones as a back-up means of offsite communications. In addition, confirm that satellite phones will function within plant structures (i.e., control room) to be considered an effective means of offsite communications.
Provide a proposed change to delete satellite phones if its use is not proceduralized, since this should be considered an extraordinary means.
GGNS Response The satellite phones are used as a means of backup communications and are tested on a periodic basis. The Emergency Plan Procedure, 10-S-01-6, Notification of Offsite Agencies and Plant On-Call Emergency Personnel, addresses the use of the satellite phone as backup communications.
- d. Describe what offsite communications capability is provided by OHL, which is listed in licensee Table S2, but not described in licensee Basis.
GGNS Response The Operational Hot Line (OHL) is a direct ringdown system for providing offsite communications with all affected State and local agencies as described in the GGNS emergency plan. When the OHL phone receiver is lifted, the phone automatically rings at all the agencies.
Licensee revised 1st NEI 99-01 criterion to state, AC power to 15AA and 16AB busses reduced to only one of the following source for > 15 minutes. If a power source is only supplying one essential bus, then a station blackout condition does not exist; yet, EAL implies that classification is warranted since both 15AA and 16AB are not being supplied. Provide proposed change to comply with NEI 99-01 guidance. In addition, identify any deviations in Attachment 5 and provide technical justification.
to GNRO-2004/00057 Page 15 of 17 GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI.
SS3 / EAL 1 (corresponds to NEI 99-01, SS2 / EAL 1)
- a. Identify deviation from NEI 99-01 wording in Attachment 5 and provide technical justification for further consideration, or provide proposed change for consistency with NEI 99-01 guidance.
GGNS Response The EAL wording was revised to be consistent with NEI and site specific indications were included.
- b. Define term rapidly insert in Attachment 4, since its use is not defined in NEI 99-01 guidance.
GGNS Response The term was deleted and the EAL revised to be consistent with EOPs.
- 40. SA5 (corresponds to NEI 99-01, CA4)
- a. Define Containment Closure in Attachment 4 consistent with definition provided in Section 5.4 to NEI 99-01. In addition, clarify why term Containment Closure is not used in licensee EALs.
GGNS Response This term was used and the definition section of the EAL implementing procedure will contain NEI 99-01 definitions when EALs are approved and procedure is issued.
Attachment 4 states, RPV pressure would have to be monitored on the Plant Data System computer to determine the 10 psig pressure increase. Per the NEI 99-01, CA4 / EAL 3 Basis, the site-specific RCS pressure chosen should be 10 psig or the lowest pressure that the site can read on installed Control Room instrumentation that is equal to or greater than 10 psig, rather than the Plant Data System computers capability. Describe the capability of Control Room instrumentation to monitor RPV pressure equal to or greater than 10 psig, and provide a proposed change, if applicable, to reflect the lowest pressure that can be read on Control Room instrumentation if minimum capability is greater than 10 psig.
GGNS Response This question is not applicable to Grand Gulf. Neither the original (SA5) or the revised (CA3) submittal state that the PDS computer would be used to monitor RPV pressure increase.
Define term significant transient in Attachment 4, consistent with that defined in Section 5.4 to NEI 99-01.
GGNS Response EAL implementing procedure will contain NEI 99-01 definitions when EALs are approved and procedure is issued.
Identify the combining loss of onsite and offsite criterion as a deviation, rather than a deviation, in Attachment 5 and provide technical justification.
to GNRO-2004/00057 Page 16 of 17 GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- 43. SS4 / EAL 1.a & b (corresponds to NEI 99-01, CS1 / EAL 2.b and CS2 / EALs 1.b and 2.b) Provide proposed change to address containment high range radiation monitor reading, erratic source range monitor indication, and other site-specific indications in compliance with NEI 99-01 guidance.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
Provide proposed change to address indication of core uncovery in Refueling Mode, versus unexplained RPV inventory loss in Cold Shutdown Mode.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
Provide proposed change to clarify that fission product barriers considered per NEI 99-01 Basis are the RCS and primary containment.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
Identify the combining loss of onsite and offsite criterion as a deviation, rather than a deviation, in Attachment 5 and provide technical justification.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
The mode applicability in Attachment 4 lists Mode 3 (Hot Shutdown), which is not consistent with Attachments 3 and 5, or NEI 99-01 guidance. Provide proposed change to address inconsistency in mode applicability.
GGNS Response The mode applicability in Attachment 4 was revised to delete Mode 3.
NEI 99-01 CG1 is not included in the GGNS EAL scheme. In explaining this, GGNS lists this as a difference and states that EALs #2 and #3 are redundant. However, these conditions are different and not addressed by the listed EALs. In a shutdown condition, significant loss of reactor coolant inventory could precede pressure, temperature, or radiological indications by an extended amount of time due to the lower decay heat conditions which would exist. In other EALs, GGNS states that RPV level indications may be suspect, or unavailable during certain plant configurations during shutdown. This EAL is intended to provide an additional indication that a significant loss of coolant has occurred which is in the process of uncovering the core. This is considered a deviation from NEI 99-01, which does not appear to be addressed with sufficient justification. Provide indications to meet this EAL, or explain this deviation with detailed justification and documentation to support that the conditions for CG1 in NEI 99-01 would be consistently classified as a General Emergency in the GGNS scheme.
to GNRO-2004/00057 Page 17 of 17 GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- 49. SG3 (corresponds to NEI 99-01, CG1)
- a. Licensee EAL criteria do not address NEI 99-01 CG1 / EAL 2 indications for high range radiation monitor readings, source range monitor readings, or other site specific indications.
The licensee lists this as a difference and uses the core uncovery criteria to satisfy this EAL.
Provide proposed change to comply with NEI 99-01 guidance, or identify as deviation in Attachment 5 and provide justification that the license EAL scheme would consistently declare a General Emergency for these conditions.
GGNS Response The EAL was revised in Attachments 3, 4, and 5 to comply with NEI 99-01.
- b. In Attachment 4, licensee applies an AND between EALs 1 and 2, indicating that both conditions must apply for the general emergency classification. NEI 99-01 lists these EALs independently, and they should stand alone as criteria as classification. Attachment 5 does not indicate that this deviation was incorporated. Provide proposed change to comply with NEI 99-01 guidance, or identify as a deviation in Attachment 5 and provide technical justification.
GGNS Response The EAL was revised in Attachment 4 to comply with NEI 99-01.
Attachment II To GNRO-2004/00057 Cross Reference Matrix
Attachment II to GNRO-2004/00057 Page 1 of 3 GGNS CROSS-REFERENCE MATRIX NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised AU1 - release > 2 X ODCM AU1 AU1 AU2 -increase in plant radiation level AU2 AU2 AA1 - release > 200 X ODCM AA1 AA1 AA2 - damage to fuel or loss of level AA3 AA2 AA3 -impedes operation AA2 AA3 AS1 - release with offsite dose AS1 AS1 AG1 - release with offsite dose AG1 AG1 NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised CU1 - RCS leakage CSD SU4 CU1 (deviation)
CU2 -loss of RCS inventory RF SU5 CU2 CU3 - LOSP >15 min SU1 CU5 CU4 - loss of decay heat removal SU6 CU3 CU5 - fuel clad degradation SU10 N/A (deviation)
CU6 - loss of communications SU9 CU8 CU7 - loss of DC power SU2 CU6 CU8 - inadvertent criticality SU7 CU7 CA1 - loss of RCS inventory CSD SA4 CA1 CA2 - loss of RCS inventory RF SA4 CA2 CA3 - SBO >15 min. SA2 CA5 CA4 - loss of decay heat removal SA5 CA3 CS1 - loss of RCS inventory CSD SS4 CS1 CS2 - loss of RCS inventory RF SS4 CS2 CG1 - loss of RCS inventory / FPBO SG3 CG1 NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised EHU1 - damage EHU1 EU1 EHU2 -security event EHU2 EU2
Attachment II to GNRO-2004/00057 Page 2 of 3 NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised Final Fuel Clad - coolant activity level Fuel Clad coolant activity level (FC1) Fuel Clad coolant activity level (FC1)
Fuel Clad - RPV level Fuel Clad RPV level (FC2) Fuel Clad RPV level (FC2)
Fuel Clad - Drywell radiation level Fuel Clad Containment rad level (FC3) Fuel Clad Containment rad level (FC3)
Fuel Clad - other site specific Fuel Clad MSL radiation monitor (FC4) Deleted Fuel Clad - ED judgment Fuel Clad ED judgment (FC5) Fuel Clad ED judgment (FC4)
RCS - drywell pressure RCS drywell pressure (RC1) RCS drywell pressure (RC1)
RCS - RPV level RCS RPV level (RC2) RCS RPV level (RC2)
RCS - RCS leak rate RCS RCS leak rate (RC3) RCS RCS leak rate (RC3)
RCS - drywell rad level N/A RCS drywell rad level (RC4)
RCS - other site specific RCS primary sys relief valves (RC4) Deleted RCS - ED judgment RCS ED judgment (RC5) RCS ED judgment (RC5)
PC - drywell pressure PC cnmt pressure (PC1) PC cnmt pressure (PC1)
PC - RPV water level PC RPV water level (PC3) PC RPV water level (PC2)
PC - cnmt isolation failure/bypass PC cnmt isolation failure/bypass (PC4) PC cnmt isolation failure/bypass (PC3)
PC - cnmt rad level PC cnmt rad level (PC5) PC cnmt rad level (PC4)
PC - other site specific PC hydrogen concentration (PC2) Deleted PC - ED judgment PC ED judgment (PC5) PC ED judgment (PC5)
NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised HU1 - natural phenomena HU3 HU6 HU2 - fire in PA HU4 HU4 HU3 - toxic or flammable gas HU5 HU5 HU4 - security event HU1 HU1 HU5 - ED judgment HU2 HU2 HA1 - natural phenomena HA4 HA6 HA2 - fire or explosion HA5 HA4 HA3 - toxic or flammable gas HA6 HA5 HA4 - security event HA1 HA1 HA5 - control room evacuation HA3 HA3 HA6 - ED judgment HA2 HA2 HS1 -.security event HS1 HS1 HS2 - control room evacuation HS3 HS3 HS3 - ED judgment HS2 HS2 HG1 - security event HG1 HG1 HG2 - ED judgment HG2 HG2
Attachment II to GNRO-2004/00057 Page 3 of 3 NEI 99-01 Rev 4 GGNS Original submittal GGNS Revised SU1 - loss of offsite power SU1 SU1 SU2 - does not reach S/D within SU11 SU11 limits SU3 - loss of annunciation SU8 SU6 SU4 - fuel clad degradation SU10 SU9 SU5 - RCS leakage SU3 SU7 SU6 - loss of onsite & offsite SU9 SU8 communications SU8 - inadvertent critically SU7 SU10 SA2 - failure of RPS SA3 SA3 SA4 - loss of annunciation and SA6 SA6 transient SA5 - single AC source SA1 SA1 SS1 - SBO SS1 SS1 SS2 - failure of RPS and manual SS3 SS3 SCRAM SS3 - loss of DC SS2 SS4 SS4 - loss of decay heat removal SS5 SS5 SS6 - inability to monitor significant SS6 SS6 transient SG1 - long term SBO SG1 SG1 SG2 - failure to SCRAM and SG2 SG3 challenge to RCS
Attachment III To GNRO-2004/00057 Proposed EALs - To Be Incorporated in Procedure
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment I Page 1 of 20 EMERGENCY CLASSIFICATIONS TABLE OF CONTENTS RECOGNITION CATEGORY PAGE
- 1. Abnormal Rad Levels / Radiological Effluents 3-4 Radiological Effluents ------------------------------------------------------------------- 3 Abnormal Rad Levels-------------------------------------------------------------------- 4
- 2. Cold Shutdown/Refueling ------------------------------------------------------------ 9 RCS Leakage ------------------------------------------------------------- 5 Loss of Decay Heat Removal --------------------------------------------------------- 7 Loss of AC Power ------------------------------------------------------------- 8 Loss of DC Power ------------------------------------------------------------- 8 Inadvertent Criticality ------------------------------------------------------------- 8 Loss of Communication ------------------------------------------------------------- 9
- 3. Events Related to Independent Spent Fuel Storage Installations -------------- 10
- 4. Fission Product Barrier Degradation --------------------------------------------------- 12 Fission Product Barrier Matrix ------------------------------------------------------- 12 Table F1 -------------------------------------------------------------------------------------- 12
- 5. Hazards and Other Malfunctions 13 - 16 Security Events---------------------------------------------------------------------------- 13 Discretionary ------------------------------------------------------------------------------- 13 Control Room Evacuation-------------------------------------------------------------- 13 Fire--------------------------------------------------------------------------------------------- 14 Toxic or Flammable Gasses----------------------------------------------------------- 14 Natural and Destructive Phenomena ----------------------------------------------- 15
- 6. System Malfunctions 17 - 20 Loss of AC Power------------------------------------------------------------------------- 17 Failure of Reactor Protection System ---------------------------------------------- 18 Loss of DC Power------------------------------------------------------------------------- 18 Loss of Decay Heat Removal --------------------------------------------------------- 19 Loss of Annunicators / Indication --------------------------------------------------- 19 Reactor Coolant System Leakage --------------------------------------------------- 19
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment I Page 2 of 20
- 6. System Malfunctions (cont.)
Loss of Communication ---------------------------------------------------------------- 20 Cladding Degradation ------------------------------------------------------------------- 20 Inadvertent Criticality-------------------------------------------------------------------- 20 Tech Spec Time Limit Exceeded----------------------------------------------------- 20 NOTE Any changes made to information in Attachment I may require changes to EPP 01-02 (EAL Matrix).
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 3 of 20 1 2 3 4 5 D ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT AG1 1 2 3 4 5 D AS1 1 2 3 4 5 D AA1 1 2 3 4 5 D AU1 1 2 3 4 5 D Offsite dose resulting from an actual or imminent Offsite dose resulting from an actual or imminent Any UNPLANNED release of gaseous or liquid Any UNPLANNED release of gaseous or liquid release of gaseous radioactivity exceeds release of gaseous radioactivity exceeds radioactivity to the environment that exceeds radioactivity to the environment that exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the 100 mR TEDE or 500 mR Thyroid CDE for the 200 times the ODCM limit 15 minutes two times the ODCM limit for 60 minutes.
actual or projected duration of the release using actual or projected duration of the release Emergency Action Level(s): (1 or 2 or 3) Emergency Action Level(s): (1 or 2 or 3) actual meteorology Emergency Action Level(s): (1 or 2 or 3)
NOTE: If monitor reading is sustained for the time NOTE: If monitor reading is sustained for the Emergency Action Level(s): (1 or 2 or 3)
NOTE: If dose assessment results are available at period indicated in the EAL and the required time period indicated in the EAL and the NOTE: If dose assessment results are available at the time of declaration, the classification assessments can not be completed within this required assessments can not be the time of declaration, the classification should be based on EAL #2 instead of EAL period, then declaration must be made based completed within this period, then should be based on EAL #2 instead of EAL #1. While necessary declarations should on the valid radiation monitor reading. declaration must be made based on the Radiological Effluents
- 1. While necessary declarations should not be delayed awaiting results, the dose valid radiation monitor reading.
not be delayed awaiting results, the dose assessment should be initiated/completed 1. VALID reading on any effluent monitor that assessment should be initiated/completed in order to determine if the classification exceeds 400 times the Hi-Hi alarm setpoint 1. VALID reading on any effluent monitor that in order to more accurately characterize the should be subsequently escalated. established by a current radioactivity discharge exceeds four times the Hi-Hi alarm setpoint nature of the release. permit for 15 minutes. established by a current radioactivity discharge
- 1. VALID reading on one or more of the radiation permit for 60 minutes.
- 1. VALID reading on one or more of the radiation monitors that exceeds or is expected to OR monitors that exceeds or is expected to exceed the reading in Table R1 SITE AREA OR
- 2. VALID reading on one or more of the radiation exceed the reading in Table R1 GENERAL EMERGENCY for 15 minutes.
monitors that exceeds the reading in Table R1 2. VALID reading on one or more of the radiation EMERGENCY for 15 minutes.
OR ALERT for 15 minutes. monitors that exceeds the reading in Table R1 OR OR UNUSUAL EVENT for 60 minutes.
- 2. Dose assessment using actual meteorology
- 2. Dose assessment using actual meteorology indicates doses > 100 mR TEDE or > 500 mR 3. Confirmed sample analyses for gaseous or liquid OR indicates doses > 1000 mR TEDE or > 5000 thyroid CDE at or beyond the site boundary. releases indicates concentrations or release rates,
- 3. Confirmed sample analyses for gaseous or mR thyroid CDE at or beyond the site with a release duration of 15 minutes, in excess of OR liquid releases indicates concentrations or boundary . 200 times the ODCM limit.
release rates, with a release duration of
- 3. Field survey results indicate closed window OR 60 minutes, in excess of 2 times the ODCM dose rates > 100 mR/hr expected to continue limit.
- 3. Field survey results indicate closed window for > one hour; or analyses of field survey dose rates > 1000 mR/hr expected to continue samples indicate thyroid CDE of 500 mR for for > one hour; or analyses of field survey one hour of inhalation, at or beyond the site samples indicate thyroid CDE of 5000 mR boundary.
for one hour of inhalation, at or beyond site boundary.
Table R1 EAL THRESHOLD Method GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Monitor ID Threshold Monitor ID Threshold Monitor ID Threshold Monitor ID Threshold Release Point Total:
OG/Radwaste Vent FHA Vent D173003 3.37E+02 Ci/sec D173002 3.37E+01 Ci/sec D173002 3.73E+00 Ci/sec D173001 3.73E-02 Ci/sec CTMT Vent Turb Bldg Vent SBGT A/B
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 4 of 20 1 2 3 4 5 D ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT AA2 1 2 3 4 5 D AU2 1 2 3 4 5 D Damage to irradiated fuel or loss of water level that has Unexpected rise in plant radiation or will result in uncovering of irradiated fuel outside the Emergency Action Level(s): (1 or 2)
Reactor Vessel
- 1. a. VALID indication of uncontrolled water Emergency Action Level(s): (1 or 2) level drop in Upper Ctmt Pools or Aux
- 1. A VALID alarm on one or more of the following: Bldg Fuel Pools or the Fuel Transfer Canal with all irradiated fuel assemblies Ctmt Vent -------------------------------- (P601-19A- remaining covered by water G9)
FH Area Vent --------------------------- (P601-19A- AND C11)
- b. Unplanned VALID Area Radiation Ctmt 208 Airlock ----------------------- (P844-1A-A1)
Monitor Alarm on any of the following:
Ctmt Fuel Hdlg Area ------------------ (P844-1A-A3)
Aux Bldg Fuel Hdlg Area ------------ (P844-1A-A4) Ctmt 208 Airlock ------------ (P844-1A-A1)
OR Ctmt Fuel Hdlg Area ------- (P844-1A-Abnormal Rad Levels
- 2. VALID indication of uncontrolled water level drop in A3)
Upper Ctmt Pools or Aux Bldg Fuel Pools or Fuel Aux Bldg Fuel Hdlg Area-- (P844-1A-Transfer Canal that will result in irradiated fuel A4) uncovering.
- 2. Unplanned VALID Area Radiation Monitor readings rise by a factor of 1000, or full scale, over alarm setpoint.
AA3 1 2 3 4 5 D Release of radioactive material or elevated radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown Emergency Action Level(s): (1 or 2)
- 1. VALID radiation reading > 15 mR/hr (General Area) in areas requiring continous occupancy to maintain plant safety functions:
Main Control Room OR
- 2. VALID radiation readings > 1 x 104 mr/hr (General Area) in areas requiring infrequent access to maintain plant safety functions:
Any ECCS Pump Rooms RCIC Pump Room
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 5 of 20 4 5 COLD SHUTDOWN/REFUELING Plant Modes: Cold Shutdown Refueling GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT CG1 4 5 CS1 4 CA1 4 CU1 4 Loss of RCS inventory affecting fuel clad integrity Loss of RCS inventory affecting core decay heat Loss of RCS inventory RCS leakage with containment challenged with irradiated fuel removal capability Emergency Action Level(s): (1 or 2) Emergency Action Level(s):
in the RPV Emergency Action Level(s): (1 or 2)
- 1. Loss of RCS inventory as indicated by RPV 1. Loss of RCS inventory and RPV level Emergency Action Level(s): (1 and 2 and 3) 1. With CONTAINMENT CLOSURE not level < -150.3 in. (Level 1). < -41.6 in. (Level 2).
- 1. Unexpected loss of RCS inventory. established: OR AND a. RCS inventory as indicated by RPV level 2. a. Unexpected loss of RCS inventory
< -156.3 in. (6 below Level 1).
- b. RPV level cannot be monitored for >15 minutes.
OR > 30 minutes with an unexpected loss of
- b. cannot be monitored with indication of core RCS inventory.
uncovery for >30 minutes as evidenced by OR one or more of the following:
- 2. With CONTAINMENT CLOSURE established:
Containment High Range Radiation Monitor reading > 100 R/hr. a. RCS inventory as indicated by RPV level Erratic Source Range Monitor indication < -167 in. (TAF)
- 3. Indication of CONTAINMENT challenged as b. RPV level cannot be monitored for indicated by one or more of the following: > 30 minutes with either:
Explosive mixture inside Containment Unexpected loss of RCS inventory.
Erratic Source Range Monitor Containment pressure >15 psig indication CONTAINMENT CLOSURE not established Secondary Containment area radiation monitor above the EOP Max Safe Operating Value in Table C1.
Table C1 Area Max Safe Operating Value RHR Room A 8 x 104 mr/hr RHR Room B 8 x 104 mr/hr RHR HX A Hatch 8 x 104 mr/hr RHR HX B Hatch 8 x 104 mr/hr RCIC Room 8 x 104 mr/hr MSL Rad Monitor 8 x 104 mr/hr
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 6 of 20 SGTS Fltr. Trn. 8 x 102 mr/hr CG1 4 5 CS2 5 CA2 5 CU2 5 Loss of RCS inventory affecting fuel clad integrity Loss of RCS inventory affecting core decay heat Loss of RCS inventory with irradiated fuel in the RPV UNPLANNED loss of RCS inventory with with containment challenged with irradiated fuel removal capability with irradiated fuel in the RPV Emergency Action Level(s): (1 or 2) irradiated fuel in the RPV in the RPV Emergency Action Level(s): (1 or 2) Emergency Action Level(s): (1 or 2)
- 1. Loss of RCS inventory as indicated by RPV level Emergency Action Level(s): (1 and 2 and 3) 1. With CONTAINMENT CLOSURE not < -150.3 in. (Level 1) 1. UNPLANNED RPV level drop below the RPV
- 1. Unexpected loss of RCS inventory. established: OR flange for 15 minutes AND a. RCS inventory as indicated by RPV level 2. a. Unexpected loss of RCS inventory. OR
< -156.3 in. (6 below Level 1)
indication of core uncovery as evidenced
- b. cannot be monitored with indication of core by one or more of the following:
uncovery for >30 minutes as evidenced by Containment High Range Radiation one or more of the following: Monitor reading > 100 R/hr.
Containment High Range Radiation Erratic Source Range Monitor Monitor reading > 100 R/hr. indication.
Erratic Source Range Monitor indication OR RCS Leakage AND
- 2. With CONTAINMENT CLOSURE established:
- 3. Indication of CONTAINMENT challenged as indicated by one or more of the following: a. RCS inventory as indicated by RPV level
< -167 in. (TAF)
Explosive mixture inside Containment OR Containment pressure >15 psig
- b. RPV level cannot be monitored with CONTAINMENT CLOSURE not indication of core uncovery as evidenced established by one or more of the following:
Secondary Containment area radiation Containment High Range Radiation monitor above the EOP Max Safe Monitor reading > 100 R/hr.
Operating Value in Table C1.
Erratic Source Range Monitor indication Table C1 Area Max Safe Operating Value RHR Room A 8 x 104 mr/hr RHR Room B 8 x 104 mr/hr RHR HX A Hatch 8 x 104 mr/hr RHR HX B Hatch 8 x 104 mr/hr RCIC Room 8 x 104 mr/hr MSL Rad Monitor 8 x 104 mr/hr SGTS Fltr. Trn. 8 x 102 mr/hr
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 7 of 20 4 5 COLD SHUTDOWN/REFUELING Plant Modes: Cold Shutdown Refueling GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT CA3 4 5 CU3 4 5 Inability to maintain plant in Cold Shutdown with UNPLANNED loss of decay heat removal irradiated fuel in the RPV capability with irradiated fuel in the RPV Emergency Action Level(s): (1 or 2 or 3) Emergency Action Level(s): (1 or 2)
Loss of Decay Heat Removal
- 1. With CONTAINMENT CLOSURE not established 1. An UNPLANNED event results in RCS and RCS integrity not established an temperature exceeding 200 °F UNPLANNED event results in RCS temperature OR exceeding 200 °F.
- 2. With CONTAINMENT CLOSURE established and RCS integrity not established an UNPLANNED event results in RCS temperature exceeding 200 °F for >20 minutes See Note OR
- 3. An UNPLANNED event results in RCS temperature exceeding 200 °F for >60 minutes See Note or results in RPV pressure rise >10 psig Note: If a heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 8 of 20 4 5 COLD SHUTDOWN/REFUELING Plant Modes: Cold Shutdown Refueling GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT CA5 4 5 D CU5 4 5 Loss of All Offsite and Loss of All Onsite AC power to Loss of all offsite power to Div I & II ESF busses Div I & II ESF busses for >15 minutes Emergency Action Level(s): Emergency Action Level(s):
- 1. a. Loss of power to all of the following 1. a. Loss of power to all of the following transformers: transformers for > 15 minutes:
Loss of AC Power ESF-11 ESF-11 ESF-21 ESF-21 ESF-12 ESF-12 AND AND
- b. Failure of both Div. I and Div. II Diesel b. At least Div. I or Div. II Diesel Generators Generators to supply power to emergency are supplying power to emergency busses. busses.
AND
- c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power CU6 4 5 UNPLANNED loss of required DC power for
>15 minutes Loss of DC Power Emergency Action Level(s):
- 1. a. UNPLANNED loss of Vital DC power to required DC busses based on < 105 VDC bus voltage indications.
AND
- b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.
CU7 4 5 Inadvertent Inadvertent criticality Emergency Action Level(s):
Criticality 1. An UNPLANNED extended positive period observed on nuclear instrumentation
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 9 of 20 4 5 COLD SHUTDOWN/REFUELING Plant Modes: Cold Shutdown Refueling GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT CU8 4 5 UNPLANNED loss of all onsite or offsite communications capabilities Emergency Action Level(s): (1 or 2)
- 1. Loss of all onsite communications capability affecting the ability to perform routine operations (See Table C2)
OR Loss of Communications
- 2. Loss of all offsite communications capability (See Table C3)
Table C2 Table C3 Offsite Communications Onsite Equipment Communications Equipment Plant Radio All telephone lines System (commercial &
fiber optic)
Plant Paging System Satellite telephone Sound Powered OHL Phones NRC phones In-plant (ENS, HPN, MCL, Telephones RSCL, PMCL)
UHF Radios
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 10 of 20 ISFSI Plant Modes: 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT E-HU1 1 2 3 4 5 D Damage to a loaded cask CONFINEMENT BOUNDARY Emergency Action Level(s): (1 or 2 or 3)
- 1. Natural phenomena events affecting a loaded cask CONFINEMENT BOUNDARY.
Tornado Confinement Boundary Damage Hurricane force winds Earthquake Flood Lightning PROPOSED OR Extreme Environmental Temperature DRAFT
- 2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.
Dropped cask Cask Tip-over Explosive Overpressure Damage by Missile generated by Natural Phenomena Fire Damage Blockage of air vents and/or burial under debris OR
- 3. Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage CONFINEMENT BOUNDARY.
E-HU2 1 2 3 4 5 D PROPOSED Confirmed Security Event with potential loss of level of safety of the ISFSI.
Security Emergency Action Level(s):
DRAFT 1. Security event as determined from the GGNS Security Plan for ISFSI and reported by the GGNS Security Shift Supervision.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 11 of 20 1 2 3 FISSION PRODUCT BARRIER DEGRADATION Plant Modes: Power Operations Startup Hot Shutdown GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT FPB Loss / Potential Loss FG1 1 2 3 FS1 1 2 3 FA1 1 2 3 FU1 1 2 3 Loss of ANY Two Barriers AND Loss or Potential Loss or Potential Loss of ANY Two Barriers ANY Loss or ANY Potential Loss of EITHER Fuel Clad ANY Loss or ANY Potential Loss of Primary Loss of Third Barrier OR RCS Containment Emergency Action Level(s):
Emergency Action Level(s): 1. Loss or Potential Loss of ANY Two Barriers Emergency Action Level(s): Emergency Action Level(s):
- 1. ANY Loss or ANY Potential Loss of EITHER Fuel 1. ANY Loss or ANY Potential Loss of Primary
- 1. Loss of ANY Two Barriers AND Loss or Clad or RCS Containment Potential Loss of Third Barrier.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 12 of 20 Fission Product Barrier Matrix Fuel Clad Reactor Coolant System Primary Containment Parameter Loss Potential Loss Parameter Loss Potential Loss Parameter Loss Potential Loss FC1 Primary RC1 Drywell Pressure > 1.39 psig with PC1 Primary Rapid unexplained loss of 15 psig and rising in Coolant Coolant activity Pressure Containment pressure following initial Primary Containment None indications of a reactor None Activity > 300 µCi/gm dose eq. I131 Pressure pressure rise.
coolant leak in the drywell OR FC2 RPV OR DW H2 concentration >9%
Water RPV water level < -192 in. RPV water level < -167 in. Pressure response not Level RPV water level < -167 in. OR RC2 RPV Water consistent with LOCA with indications of a reactor None conditions. Ctmt H2 concentration in FC3 Drywell Level coolant leak in the drywell HDOL Unsafe Zone Radiation Drywell Radiation monitor None Monitoring reading >3000 R/hr FC4 ED Any condition in the opinion Any condition in the RC3 Reactor Unisolable MSL break as Reactor Coolant System Judgment of the Emergency Director opinion of the Emergency Coolant indicated by the failure of both leakage > 50 gpm inside Primary Containment that indicates a loss of the Director that indicates a System MSIVs in any one line to close the drywell. PC2 RPV Water flooding required by:
None Fuel Clad Barrier. potential loss of the Fuel Leak Rate Level AND OR SAP 1, 3, 4, 5 or 6 Clad Barrier.
High MSL Flow and High Unisolable primary Steam Tunnel Temperature system leakage outside PC3 Primary Failure to isolate any Table F1 annunciators Primary Containment as Containment penetration AND indicated by any Area Isolation downstream pathway to Area Temperature OR Temperature or Area Failure or the environment exists.
Parameter Alert Limit SAE/GE Limit Direct report of steam release. Radiation > Alert Value in Bypass Table F1. OR 185°F 235°F MSL Pipe Tunnel Temp. Intentional venting per (P601-19A/18A-A3/A4) (E31-N064A,B,C,D,E,F)
165°F 225°F RHR-A Equip Area Temp. OR (P601-20A-B1) (E31-N068A,N610A) 165°F 225°F Unisolable primary None RHR-B Equip Area Temp.
(P601-20A-B1) (E31-N068B,N610B) RC4 Drywell Drywell Radiation monitor system leak outside 185°F 212°F Radiation reading > 100 R/hr with Primary Containment as RCIC Equip Area Temp. None (P601-21A-G3) (E31-N602A/B) Monitoring indications of a leak in the indicated by any Area 170°F drywell Temperature or Area RWCU Pmp Rm 1 Temp NA (P680-11A-A1) Radiation level 170°F RC5 ED Any condition in the opinion of Any condition in the > SAE / GE Value in RWCU Pmp Rm 2 Temp NA (P680-11A-A2) Judgment the Emergency Director that opinion of the Table F1.
indicates a loss of the RCS Emergency Director that Barrier. indicates a potential loss of the RCS Barrier.
Area Radiation Level Parameter Alert Limit SAE/GE Limit 102 MR/HR RHR Room A Rad 8 x 104 MR/HR (P844-1A-D4) PC4 Significant 102 MR/HR 4 Radioactive Containment Radiation RHR Room B Rad 8 x 10 MR/HR None monitor reading (P844-1A-D4) Inventory in 102 MR/HR Primary >10,000 R/hr RHR HX A Hatch Rad 8 x 104 MR/HR Containment (P844-1A-C4) 102 MR/HR RHR HX B Hatch Rad 8 x 104 MR/HR (P844-1A-C4) 102 MR/HR PC5 ED Judgment Any condition in the Any condition in the opinion RCIC Room Rad 8 x 104 MR/HR (P844-1A-D4) opinion of the Emergency of the Emergency Director Director that indicates a that indicates a potential loss of the Primary loss of the Primary Containment barrier. Containment barrier.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 13 of 20 1 2 3 4 5 D HAZARDS AND OTHER MALFUNCTIONS Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HG1 1 2 3 4 5 D HS1 1 2 3 4 5 D HA1 1 2 3 4 5 D HU1 1 2 3 4 5 D Security event resulting in loss of physical control of Confirmed security event in a plant VITAL AREA Confirmed security event in a plant Confirmed security event which indicates a potential the facility PROTECTED AREA degradation in the level of safety of the plant Security Events Emergency Action Level(s): (1 or 2)
Emergency Action Level(s): Emergency Action Level(s): (1 or 2) Emergency Action Level(s): (1 or 2)
- 1. INTRUSION into the plant VITAL AREA by a
- 1. A HOSTILE FORCE has taken control of plant HOSTILE FORCE. 1. INTRUSION into the plant PROTECTED AREA 1. Security events as determined from the equipment such that plant personnel are by a HOSTILE FORCE. GGNS Safeguards Contingency Plan and OR unable to operate equipment required to reported by the GGNS security shift OR maintain safety functions. 2. Other security events as determined from the supervision.
GGNS Safeguards Contingency Plan and 2. Other security events as determined from the OR reported by the GGNS security shift GGNS Safeguards Contingency Plan and supervision. reported by the GGNS security shift supervision 2. A credible site specific security threat notification.
HG2 1 2 3 4 5 D HS2 1 2 3 4 5 D HA2 1 2 3 4 5 D HU2 1 2 3 4 5 D Other Conditions existing which in the judgment of Other Conditions existing which in the judgment of Other Conditions existing which in the judgment of the Other Conditions existing which in the judgment of the Emergency Director warrant declaration of the Emergency Director warrant declaration of Emergency Director warrant declaration of an ALERT the Emergency Director warrant declaration of an GENERAL EMERGENCY SITE AREA EMERGENCY UNUSUAL EVENT Emergency Action Level(s):
Discretionary Emergency Action Level(s): Emergency Action Level(s): Emergency Action Level(s):
- 1. Other conditions exist which in the judgment of
- 1. Other conditions exist which in the judgment 1. Other conditions exist which in the judgment the Emergency Director indicate that events are 1. Other conditions exist which in the of the Emergency Director indicate that of the Emergency Director indicate that in progress or have occurred which involve judgment of the Emergency Director events are in progress or have occurred events are in progress or have occurred actual or likely potential substantial degradation indicate that events are in progress or have which involve actual or imminent substantial which involve actual or likely major failures of the level of safety of the plant. Any releases occurred which indicate a potential core degradation or melting with potential for of plant functions needed for protection of are expected to be limited to small fractions of degradation of the level of safety of the loss of containment integrity. Releases can the public. Any releases are not expected to the EPA Protective Action Guideline exposure plant. No releases of radioactive material be reasonably expected to exceed EPA result in exposure levels which exceed EPA levels. requiring offsite response or monitoring Protective Action Guideline exposure levels Protective Action Guideline exposure levels are expected unless further degradation of offsite for more than the immediate site area. beyond the site boundary. safety systems occurs.
HS3 1 2 3 4 5 D HA3 1 2 3 4 5 D Control Room evacuation has been Initiated and Control Room evacuation has been initiated Control Room plant control cannot be established Emergency Action Level(s):
Emergency Action Level(s):
- 1. Entry into 05-1-02-II-1, Shutdown from the Remote
- 1. Control Room evacuation has been initiated Shutdown Panel, for Control Room evacuation.
Evacuation AND Control of the plant cannot be established per 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, within 15 minutes
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 14 of 20 1 2 3 4 5 D HAZARDS AND OTHER MALFUNCTIONS Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HA4 1 2 3 4 5 D HU4 1 2 3 4 5 D FIRE or EXPLOSION affecting the operability of plant FIRE within PROTECTED AREA boundary not safety systems required to establish or maintain safe extinguished within 15 minutes of detection shutdown Emergency Action Level(s):
Emergency Action Level(s):
- 1. FIRE in Buildings or areas Contiguous to
- 1. FIRE or EXPLOSION in Structures containing any VITAL AREA (Table H3) not Functions or Systems Required for Safe extinguished within 15 minutes of control Fire Shutdown (Table H2) room notification or verification of a control room alarm.
AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified areas.
HA5 1 2 3 4 5 D HU5 1 2 3 4 5 D Release of toxic or flammable gases within or Release of toxic or flammable gases deemed contiguous to a VITAL AREA which jeopardizes detrimental to Normal Operation of the plant Toxic or Flammable Gases operation of systems required to maintain safe operations or establish or maintain safe shutdown. Emergency Action Level(s): (1 or 2)
Emergency Action Level(s): (1 or 2)
- 1. Report or detection of toxic or flammable gases that has or could enter the site
- 1. Report or detection of toxic gases within Buildings area boundary in amounts that can affect or areas Contiguous to a VITAL AREA (Table H3) NORMAL PLANT OPERATIONS.
in concentrations that may result in an atmosphere OR IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH) 2. Report by Local, County/Parish or State Officials for evacuation or sheltering of site OR personnel based on an offsite event.
- 2. Report or detection of gases in concentrations greater than the LOWER FLAMMABILITY LIMIT within Buildings or areas Contiguous to a VITAL AREA (Table H3)
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 15 of 20 1 2 3 4 5 D HAZARDS AND OTHER MALFUNCTIONS Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HA6 1 2 3 4 5 D HU6 1 2 3 4 5 D Natural and destructive phenomena affecting the Plant Natural and destructive phenomena affecting the VITAL AREA PROTECTED AREA Emergency Action Level(s): (1 or 2 or 3 or 4 Emergency Action Level(s): (1 or 2 or 3 or 4 or 5 or 6) or 5 or 6 or 7)
- 1. Valid indication of a seismic event greater than 1. Valid indication of a felt earthquake:
Operating Basis Earthquake (OBE):
Vibratory ground motion felt in the Receipt of all of the following indications on PROTECTED AREA and recognized as an SH13P856: earthquake.
Containment Operating Basis Earthquake AND (P856-1A-A3)
Activated seismic switches as indicated by Drywell Operating Basis Earthquake activation of the Seismic Monitoring (P856-1A-A5)
Natural and Destructive Phenomena System: Strong Motion Accelerometer Containment Safe Shutdown Earthquake System Activation (P856-1A-A1)
(P856-1A-A2)
Drywell Safe Shutdown Earthquake OR (P856-1A-A4) 2. Report by plant personnel of tornado striking Strong Motion Accelerometer System within PROTECTED AREA boundary.
Activation (P856-1A-A1)
AND OR Start of all five SMA cassette tape recorders 3. Vehicle crash into Plant Structures containing Functions or Systems Required for Safe AND Shutdown (Table H2) within PROTECTED Event indicator flag changes from black to white AREA boundary.
AND OR Event alarm yellow light in ON 4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA OR boundary resulting in VISIBLE DAMAGE to
- 2. Tornado striking within PROTECTED AREA permanent structure or equipment.
boundary and resulting in VISIBLE DAMAGE to any OR of the Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2) or 5. Report of turbine failure resulting in casing Control Room indication of degraded performance penetration or damage to turbine or generator of those systems seals OR OR
- 3. Vehicle crash within PROTECTED AREA 6. Uncontrolled flooding in the Auxiliary Building boundary and resulting in VISIBLE DAMAGE to (Table H1) that has the potential to affect any of the Plant Structures containing Functions or safety related equipment needed for the Systems Required for Safe Shutdown (Table H2) current operating mode.
or Control Room indication of degraded OR (continued on next page) performance of those systems.
OR (continued on next page)
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 16 of 20 1 2 3 4 5 D HAZARDS AND OTHER MALFUNCTIONS Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT OR (continued from previous page) OR (continued from previous page)
- 4. Turbine failure-generated missiles result in any 7. Severe weather with indication of sustained VISIBLE DAMAGE to or penetration of any of the high winds 74 mph within PROTECTED Natural and Destructive Phenomena (cont.)
Plant Structures containing Functions or Systems AREA boundary Required for Safe Shutdown (Table H2).
- 5. Uncontrolled flooding in the Auxiliary Building (Table H1) that results in degraded safety system performance as indicated in the Control Room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.
- 6. Severe weather with indication of sustained high winds 74 mph within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2) or has caused damage as evidenced by Control Room indication of degraded performance of those systems Table H1 Table H2 Table H3 Auxiliary Building Area Parameters Structures Containing Functions or Systems Buildings or Areas Contiguous Required for Safe Shutdown To Any VITAL AREAS Area Max Safe Operating Value Tables RHR Room A 93 FT. 6 IN. (P870-2A-E1) Unit I Containment Unit I Containment RHR Room B 93 FT. 6 IN. (P870-10A-G1) Unit I Auxiliary Building Unit I & II Auxiliary Building RHR Room C 93 FT. 6 IN. (P870-10A-G2) Control Building Control Building RCIC Room 93 FT. 6 IN. (P870-2A-A1) Unit 1 Turbine Building Unit I & II Turbine Building LPCS Room 93 FT. 6 IN. (P870-2A-F1) Diesel Generator Rooms Diesel Generator Rooms HPCS Room 93 FT. 6 IN. (P870-5A-H1) SSW Pump & Valve Rooms SSW Pump & Valve Rooms
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 17 of 20 1 2 3 4 5 D SYSTEM MALFUNCTIONS Plant Modes: Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SG1 1 2 3 SS1 1 2 3 SA1 1 2 3 SU1 1 2 3 Prolonged loss of all offsite power and prolonged Loss of all offsite power and loss of all onsite AC AC power capability to Div I or II ESF busses reduced Loss of all offsite power to Div I & II ESF busses loss of all onsite AC power to Div I, II & III ESF power to Div I, II & III ESF busses to a single power source for >15 minutes such that any for >15 minutes busses additional single failure would result in Emergency Action Level(s): Emergency Action Level(s):
Station Blackout.
Emergency Action Level(s):
- 1. Loss of power to all of the following 1. Loss of power to all of the following Emergency Action Level(s):
- 1. Loss of power to all of the following transformers: transformers for > 15 minutes:
transformers: 1. AC power capability to 15AA or 16AB reduced to ESF-11 ESF-11 Loss of AC Power a single power source for >15 minutes.
ESF-11 ESF-21 ESF-21 AND ESF-21 ESF-12 ESF-12 ESF-12 Any additional single failure will result in AND AND Station Blackout.
AND Failure of Div. I, II and III Diesel Generators to At least Div. I and Div. II Diesel Generators Failure of Div. I, II and III Diesel Generators to supply power to emergency busses. are supplying power to emergency busses.
supply power to emergency busses.
AND AND Failure to restore power to at least one Either of the following: (a or b) emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power
- a. Restoration of at least one emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely OR
- b. RPV level can not be maintained > -192 in.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 18 of 20 SYSTEM MALFUNCTIONS Plant Modes: 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SG3 1 2 SS3 1 2 SA3 1 2 Failure of the Reactor Protection System to Failure of Reactor Protection System Failure of Reactor Protection System instrumentation to complete an automatic scram and manual scram instrumentation to complete or initiate an automatic complete or initiate an automatic reactor scram once a was NOT successful and there is indication of an reactor scram once a Reactor Protection System Reactor Protection System setpoint has been extreme challenge to the ability to cool the core setpoint has been exceeded and manual scram exceeded and manual scram was successful.
Failure of Reactor Protection System was NOT successful Emergency Action Level(s): Emergency Action Level(s):
Emergency Action Level(s):
- 1. Indication(s) exist that indicate that Reactor 1. Indication(s) exist that indicate that Reactor Protection System (RPS) setpoint was 1. Indication(s) exist that indicate that Reactor Protection System (RPS) setpoint was exceeded exceeded and Protection System (RPS) setpoint was and exceeded and RPS automatic scram and a manual scram RPS automatic scram was unsuccessful as or ARI fails to reduce reactor power to RPS automatic scram and a manual scram evidenced by:
< 4%. or ARI fails to reduce reactor power to 3/4 More than one control rod at position
< 4%,
AND 02 or greater, Either of the following: (a or b) OR
- a. Indication(s) exists that the core 3/4 Control rod position is unknown for cooling is extremely challenged. more than one control rod and SRMs are either upscale or count rate is Entry into SAPs rising.
OR AND
- b. Indication(s) exists that heat removal Manual scram or ARI reduces reactor power to is extremely challenged.
< 4%.
RPV pressure and suppression pool temperature cannot be maintained in the Heat Capacity Temperature Limit (HCTL) Safe Zone.
SS4 1 2 3 Loss of all vital DC power Loss of DC Emergency Action Level(s):
Power
- 1. Loss of all Vital DC power based on < 105 VDC bus voltage indications for >15 minutes.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 19 of 20 SYSTEM MALFUNCTIONS Plant Modes: 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SS5 1 2 3 Loss of Decay Complete loss of heat removal capability Emergency Action Level(s):
Heat Removal
- 1. Heat Capacity Temperature Limit Curve exceeded.
SS6 1 2 3 SA6 1 2 3 SU6 1 2 3 Inability to monitor a SIGNIFICANT TRANSIENT in UNPLANNED loss of most or all safety system UNPLANNED loss of most or all safety system progress annunciation or indication in control room with either annunciation or indication in the control room for Emergency Action Level(s): (1) a SIGNIFICANT TRANSIENT in progress, or > 15 minutes Loss of Annunciors / Indication (2) compensatory non-alarming indicators are
- 1. a. Loss of most or all Control Room Emergency Action Level(s):
unavailable annunciators associated with safety 1. UNPLANNED loss of most or all Control systems. Emergency Action Level(s): Room annunciators or indicators associated AND 1. UNPLANNED loss of most or all Control Room with safety systems for > 15 minutes.
annunciators or indicators associated with safety
- b. Compensatory non-alarming indications systems for > 15 minutes.
are unavailable.
AND AND Either of the following: (a or b)
- c. Indications needed to monitor safety functions are unavailable. a. A SIGNIFICANT TRANSIENT is in progress.
Reactivity control OR Core cooling b. Compensatory non-alarming indications are RCS status unavailable.
CONTAINMENT status AND
- d. SIGNIFICANT TRANSIENT in progress.
SU7 1 2 3 Reactor Coolant RCS Leakage Emergency Action Level(s): (1 or 2)
- 1. Unidentified or pressure boundary leakage
> 10 gpm.
System Leakage OR
- 2. Identified leakage > 35 gpm.
Grand Gulf Nuclear Station Emergency Plan Procedure 10-S-01-1 Revision xx Attachment 1 Page 20 of 20 SYSTEM MALFUNCTIONS Plant Modes: 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled UNUSUAL EVENT SU8 1 2 3 Loss of Communication Table S1 Table S2 UNPLANNED loss of all onsite or offsite Onsite Communications Equipment Offsite Communications Equipment communications capabilities Plant Radio System All telephone lines (commercial & fiber optic) Emergency Action Level(s): (1 or 2)
Plant Paging System Satellite telephone 1. Loss of all onsite communications capability affecting the ability to perform routine Sound Powered Phones OHL operations.
In-plant Telephones NRC phones (ENS, HPN, MCL, RSCL, PMCL) (See Table S1)
- 2. Loss of all offsite communications capability.
(See Table S2)
SU9 1 2 3 Cladding Degradation Fuel clad degradation Emergency Action Level(s): (1 or 2)
- 1. Offgas isolation (1H13-P601-19A-C8) due to valid Offgas Post Treatment monitor signal indicating fuel clad degradation.
- 2. Reactor coolant sample activity > 4.0 µCi/gm dose equivalent I131, indicating fuel clad degradation.
SU10 3 Inadvertent Inadvertent criticality Emergency Action Level(s):
Criticality
- 1. An UNPLANNED extended positive period observed on nuclear instrumentation SU11 1 2 3 Tech Spec Time Inability to reach required shutdown within Technical Specification limits Emergency Action Level(s):
Limit Exceeded
- 1. Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time
Attachment IV To GNRO-2004/00057 Proposed EAL Bases - To Be Incorporated in Procedure
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 1 of 110 BASES for EMERGENCY CLASSIFICATIONS TABLE OF CONTENTS RECOGNITION CATEGORY PAGE
- 1. Abnormal Rad Levels / Radiological Effluents 4 - 16 AU1 --------------------------------------------------------------------------------------------- 4 AU2 --------------------------------------------------------------------------------------------- 6 AA1 --------------------------------------------------------------------------------------------- 8 AA2 --------------------------------------------------------------------------------------------- 10 AA3 --------------------------------------------------------------------------------------------- 11 AS1---------------------------------------------------------------------------------------------- 13 AG1 --------------------------------------------------------------------------------------------- 15
- 2. Cold Shutdown/Refueling 17 - 40 CU1 --------------------------------------------------------------------------------------------- 17 CU2 --------------------------------------------------------------------------------------------- 18 CU3 --------------------------------------------------------------------------------------------- 20 CU5 --------------------------------------------------------------------------------------------- 22 CU6 --------------------------------------------------------------------------------------------- 23 CU7 --------------------------------------------------------------------------------------------- 24 CU8 --------------------------------------------------------------------------------------------- 25 CA1 --------------------------------------------------------------------------------------------- 27 CA2 --------------------------------------------------------------------------------------------- 29 CA3 --------------------------------------------------------------------------------------------- 31 CA5 --------------------------------------------------------------------------------------------- 33 CS1---------------------------------------------------------------------------------------------- 34 CS2---------------------------------------------------------------------------------------------- 36 CG1 --------------------------------------------------------------------------------------------- 38
- 3. Events Related to Independent Spent Fuel Storage Installations 41 - 43 E-HU1 ------------------------------------------------------------------------------------------ 41 E-HU2 ------------------------------------------------------------------------------------------ 43
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 2 of 110
- 4. Fission Product Barrier Degradation ---------------------------------------------------- 44 - 62 FU1---------------------------------------------------------------------------------------------- 44 FA1---------------------------------------------------------------------------------------------- 45 FS1 ---------------------------------------------------------------------------------------------- 46 FG1---------------------------------------------------------------------------------------------- 47 Loss or Potential Loss of Fuel Clad Parameters-------------------------------- 51 FC1 - Primary Coolant Activity ----------------------------------------------------- 48 FC2 - RPV Water Level-------------------------------------------------------------- 49 FC3 - Drywell Radiation Monitoring----------------------------------------------- 50 FC4 - Emergency Director Judgment -------------------------------------------- 51 Loss or Potential Loss of Reactor Coolant System Parameters----------- 57 RC1 - Drywell Pressure-------------------------------------------------------------- 52 RC2 - RPV Water Level ------------------------------------------------------------- 53 RC3 - Reactor Coolant System Leak Rate ------------------------------------- 54 RC4 - Drywell Radiation Monitoring ---------------------------------------------- 56 RC5 - Emergency Director Judgment -------------------------------------------- 57 Loss or Potential Loss of Primary Containment Parameters --------------- 62 PC1 - Primary Containment Pressure-------------------------------------------- 58 PC2 - RPV Water Level ------------------------------------------------------------- 59 PC3 - Primary Containment Isolation Failure or Bypass -------------------- 60 PC4 - Significant Radioactive Inventory in Primary Containment --------- 61 PC5 - Emergency Director Judgment -------------------------------------------- 62
- 5. Hazards and Other Malfunctions--------------------------------------------------------- 87 HU1 --------------------------------------------------------------------------------------------- 63 HU2 --------------------------------------------------------------------------------------------- 65 HU4 --------------------------------------------------------------------------------------------- 66 HU5 --------------------------------------------------------------------------------------------- 68 HU6 --------------------------------------------------------------------------------------------- 69 HA1 --------------------------------------------------------------------------------------------- 72 HA2 --------------------------------------------------------------------------------------------- 74 HA3 --------------------------------------------------------------------------------------------- 75 HA4 --------------------------------------------------------------------------------------------- 76 HA5 --------------------------------------------------------------------------------------------- 78 HA6 --------------------------------------------------------------------------------------------- 80 HS1---------------------------------------------------------------------------------------------- 83 HS2---------------------------------------------------------------------------------------------- 84
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 3 of 110
- 5. Hazards and Other Malfunctions (cont.)
HS3---------------------------------------------------------------------------------------------- 85 HG1 --------------------------------------------------------------------------------------------- 86 HG2 --------------------------------------------------------------------------------------------- 87
- 6. System Malfunctions 88 - 110 SU1---------------------------------------------------------------------------------------------- 88 SU6---------------------------------------------------------------------------------------------- 89 SU7---------------------------------------------------------------------------------------------- 91 SU8---------------------------------------------------------------------------------------------- 92 SU9---------------------------------------------------------------------------------------------- 94 SU10 -------------------------------------------------------------------------------------------- 95 SU11 -------------------------------------------------------------------------------------------- 96 SA1---------------------------------------------------------------------------------------------- 97 SA3---------------------------------------------------------------------------------------------- 98 SA6---------------------------------------------------------------------------------------------- 99 SS1---------------------------------------------------------------------------------------------- 101 SS3---------------------------------------------------------------------------------------------- 102 SS4---------------------------------------------------------------------------------------------- 103 SS5---------------------------------------------------------------------------------------------- 104 SS6---------------------------------------------------------------------------------------------- 105 SG1 --------------------------------------------------------------------------------------------- 107 SG3 --------------------------------------------------------------------------------------------- 109
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 4 of 110 AU1 Initiating Condition -- UNUSUAL EVENT Any UNPLANNED release of gaseous or liquid radioactivity to the environment that exceeds 2 times the ODCM limit for 60 minutes Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3)
NOTE: If monitor reading is sustained for the time period indicated in the EAL AND the required assessments can not be completed within this period, then declaration must be made based on the valid radiation monitor reading.
- 1. VALID reading on any effluent monitor that exceeds four times the Hi-Hi alarm setpoint established by a current radioactivity discharge permit for 60 minutes.
- 2. VALID reading on one or more of the radiation monitors that exceeds the reading in Table R1 UNUSUAL EVENT for 60 minutes.
- 3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 60 minutes, in excess of 2 times the ODCM limit.
Table R1 EAL THRESHOLD Method UNUSUAL EVENT Monitor ID Threshold Release Point Total:
OG/Radwaste Vent FHA Vent D173001 3.73E-02 Ci/sec CTMT Vent Turb Bldg Vent SBGT A/B
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 5 of 110 AU1(cont)
Basis:
This IC addresses a potential or actual reduction in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the Offsite Dose Calculation Manual (ODCM).
The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.
The ODCM multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose or dose rate. Releases should not be prorated or averaged. For example, a release exceeding 8 x TRM Spec limit for 30 minutes does not meet the threshold for this IC.
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm set points, etc.) on the applicable permit. The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.
EAL #1 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed two times the TRM Spec. limit and releases are not terminated within 60 minutes. The alarm setpoint associated with a planned batch liquid release is established by the ODCM to alarm at 50%
of the TRM Spec. limit providing an early warning of a potential release. Upon receipt of the liquid radiation monitor alarm Chemistry is required to sample and perform an analysis. Therefore, an analysis is performed to ascertain whether or not ODCM limits have been exceeded. Indexing the EAL threshold to the ODCM setpoints in this manner insures that the EAL threshold will never be less than the setpoint established by a specific discharge permit.
EAL #2 is intended to for licensees that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally by prepared. The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments.
EAL #3 addresses uncontrolled releases that are detected by sample analysis, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage into river water systems, etc.
EALs #1 and #2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the alarm setpoints.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 6 of 110 AU2 Initiating Condition -- UNUSUAL EVENT Unexpected rise in plant radiation Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. a. VALID indication of uncontrolled water level drop in Upper Ctmt Pools or Aux Bldg Fuel Pools or the Fuel Transfer Canal with all irradiated fuel assemblies remaining covered by water AND
- b. Unplanned VALID Area Radiation Monitor Alarm on any of the following:
Ctmt 208 Airlock -----------------------------(P844-1A-A1)
Ctmt Fuel Hdlg Area------------------------(P844-1A-A3)
Aux Bldg Fuel Hdlg Area ------------------(P844-1A-A4)
- 2. Unplanned VALID Area Radiation Monitor readings rise by a factor of 1000, or full scale, over alarm setpoint.
Basis:
This IC addresses rising radiation levels as a result of a water level drop above the RPV flange or events that have resulted, or may result, in unexpected rise in radiation dose rates within plant buildings. These radiation rises represent a loss of control over radioactive material and may represent a potential degradation in the level of safety of the plant.
In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types of events via EAL #1 is appropriate given their potential for elevated doses to plant staff. Classification as an UNUSUAL EVENT is warranted as a precursor to a more serious event. Indications include local area radiation monitors and personnel reports (e.g., refueling crew). If available, video cameras may allow remote observation.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 7 of 110 AU2(cont)
While a radiation monitor could detect a rise in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. For example, the reading on an area radiation monitor located on the refueling bridge may rise due to planned evolutions such as head lift, or even a fuel assembly being raised. Generally, elevated radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss. For refueling events where the water level drops below the RPV flange classification would be via CU2, UNPLANNED loss of RCS inventory with irradiated fuel in the RPV. This event escalates to an ALERT per IC AA2 if irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier Matrix for events in Modes 1-3.
EAL #2 addresses UNPLANNED rises in in-plant radiation levels that represent a degradation in the control of radioactive material, and represent a potential degradation in the level of safety of the plant.
The term Full scale was included in the EAL to address those areas where the ARM range will not allow the determination of a rise by a factor of 1000 (i.e., the ARM range is too small). This event escalates to an ALERT per IC AA3 if the rise in dose rates impedes personnel access necessary for safe operation.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 8 of 110 AA1 Initiating Condition -- ALERT Any UNPLANNED release of gaseous or liquid radioactivity to the environment that exceeds 200 times the ODCM limit for 15 minutes Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3)
NOTE: If monitor reading is sustained for the time period indicated in the EAL and the required assessments can not be completed within this period, then declaration must be made based on the valid radiation monitor reading.
- 1. VALID reading on any effluent monitor that exceeds 400 times the Hi-Hi alarm setpoint established by a current radioactivity discharge permit for 15 minutes.
- 2. VALID reading on one or more of the radiation monitors that exceeds the reading in Table R1 ALERT for 15 minutes OR
- 3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 15 minutes, in excess of 200 times the ODCM limit.
Table R1 EAL THRESHOLD ALERT Method Monitor ID Threshold Release Points Total:
OG/Radwaste Vent FHA Vent D173002 3.73E+00 Ci/sec CTMT Vent Turb Bldg Vent SBGT A/B
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 9 of 110 AA1(cont)
Basis:
This IC addresses a potential or actual reduction in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the Offsite Dose Calculation Manual (ODCM).
The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.
The multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose or dose rate. Releases should not be prorated or averaged.
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoint, etc.) on the applicable permit. The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.
EAL #1 addresses radioactivity releases that for whatever reason cause effluent radiation monitor readings that exceed four hundred times the alarm setpoint established by the radioactivity discharge permit. This alarm setpoint is associated with a planned batch release. The setpoint is established by the ODCM to warn of a release that is not in compliance. Indexing the EAL threshold to the ODCM setpoint in this manner insures that the EAL threshold will never be less than the setpoint established by a specific discharge permit.
EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on non-routine release pathways (i.e., for which a discharge permit would not normally be prepared). The ODCM establishes a methodology for determining effluent radiation monitor setpoint. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments.
EAL #3 addresses uncontrolled releases that are detected by sample analysis, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage into river water systems, etc.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 10 of 110 AA2 Initiating Condition -- ALERT Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. A VALID alarm on one or more of the following:
Ctmt Vent------------------------------- (P601-19A-G9)
FH Area Vent-------------------------- (P601-19A-C11)
Ctmt 209 Airlock ---------------------- (P844-1A-A1)
Ctmt Fuel Hdlg Area----------------- (P844-1A-A3)
Aux Bldg Fuel Hdlg Area ----------- (P844-1A-A4)
- 2. VALID indication of uncontrolled water level drop in Upper Ctmt Pools or Aux Bldg Fuel Pools or Fuel Transfer Canal that has or will result in irradiated fuel uncovering.
Basis:
This IC addresses specific events that have resulted, or may result, in unexpected rises in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent a degradation in the level of safety of the plant. These events escalate from IC AU2 in that fuel activity has been released, or is anticipated due to fuel heat up. This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in IC E-HU1.
EAL #1 addresses elevated radiation monitor indications coincident with a potential fuel damaging event.
Elevated readings on ventilation monitors may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Application of this Initiating Condition requires understanding of the actual radiological conditions present in the vicinity of the monitor. For example, the monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head EAL #2 indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. If available, video cameras may allow remote observation.
Escalation would occur via IC AS1 or AG1 or Emergency Director judgment.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 11 of 110 AA3 Initiating Condition -- ALERT Release of radioactive material or elevated radiation levels within the facility that impedes operation of systems required to maintain safe operations or to establish or maintain cold shutdown Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. VALID radiation reading > 15 mR/hr (General Area) in areas requiring continous occupancy to maintain plant safety functions:
Main Control Room OR
- 2. VALID radiation readings > 1 x 104 mr/hr (General Area) in areas requiring infrequent access to maintain plant safety functions:
RHR A Room RHR B Room RHR C Room HPCS Room LPCS Room RCIC Room Basis:
This IC addresses elevated radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually or that requires local monitoring, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the rise in radiation levels is not a concern of this IC. The Emergency Director must consider the source or cause of the elevated radiation levels and determine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the control room may be a problem in itself. However, the rise may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission product barrier matrix ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 12 of 110 AA3 (cont.)
This IC is not meant to apply to anticipated temporary rises due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.)
The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.
For areas requiring infrequent access, the value should be based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will impede necessary access. As used here, impede, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 13 of 110 AS1 Initiating Condition -- SITE AREA EMERGENCY Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mR TEDE or 500 mR CDE Thyroid for the actual or projected duration of the release.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3)
NOTE: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated/completed in order to determine if the classification should be subsequently escalated.
- 1. VALID reading on one or more of the radiation monitors that exceeds or is expected to exceed the reading in Table R1 SITE AREA EMERGENCY for 15 minutes.
- 2. Dose assessment using actual meteorology indicates doses > 100 mR TEDE or > 500 mR thyroid CDE at or beyond the site boundary.
- 3. Field survey results indicate closed window dose rates > 100 mR/hr expected to continue for > one hour; or analyses of field survey samples indicate thyroid CDE of 500 mR for one hour of inhalation, at or beyond the site boundary.
Table R1 EAL THRESHOLD Method SITE AREA EMERGENCY Monitor ID Threshold Release Point Total:
OG/Radwaste Vent FHA Vent D173002 3.37E+01 Ci/sec CTMT Vent Turb Bldg Vent SBGT A/B
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 14 of 110 AS1 (cont.)
Basis:
This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed a small fraction of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. While these failures are addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone, e.g., fuel handling accident in spent fuel building.
The TEDE and CDE thyroid doses are set at 10% of the EPA PAG while the 500 mR thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.
Since dose assessment is based on actual meteorology, whereas the monitor readings EALs are not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures should call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EALs. Classification should not be delayed pending results of these dose assessments.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 15 of 110 AG1 Initiating Condition -- GENERAL EMERGENCY Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR CDE Thyroid for the actual or projected duration of the release using actual meteorology.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3)
NOTE: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated/completed in order to more accurately characterize the nature of the release.
- 1. VALID reading on one or more of the radiation monitors that exceeds or is expected to exceed the reading in Table R1 GENERAL EMERGENCY for 15 minutes.
- 2. Dose assessment using actual meteorology indicates doses > 1000 mR TEDE or > 5000 mR thyroid CDE at or beyond the site boundary.
OR 3 Field survey results indicate closed window dose rates > 1000 mR/hr expected to continue for > one hour; or analyses of field survey samples indicate thyroid CDE of 5000 mR for one hour of inhalation, at or beyond site boundary.
Table R1 EAL THRESHOLD Method GENERAL EMERGENCY Monitor ID Threshold Release Point Total:
OG/Radwaste Vent FHA Vent D173003 3.37E+02 Ci/sec CTMT Vent Turb Bldg Vent SBGT A/B
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 16 of 110 AG1 (cont.)
Basis:
This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. While these failures are addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that, for the more severe accidents, the release may be unmonitored or there may be large uncertainties associated with the source term and/or meteorology.
Since dose assessment is based on actual meteorology, whereas the monitor readings EALs are not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures should call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EALs. Classification should not be delayed pending results of these dose assessments.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 17 of 110 CU1 Initiating Condition -- UNUSUAL EVENT RCS leakage Operating Mode Applicability: Mode 4..............Cold Shutdown Emergency Action Level(s):
Basis:
This IC is included because it may be a precursor of more serious conditions and is considered to be a potential degradation of the level of safety of the plant. The word unplanned is used to ensure planned evolutions which reduce RPV water level below -41.6 in. do not meet the criteria of this EAL.
Prolonged loss of RCS inventory may result in escalation to the Alert level via either IC CA1 (Loss of RCS inventory) or CA3 (Inability to maintain plant in Cold Shutdown with irradiated fuel in the RPV).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 18 of 110 CU2 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of RCS inventory with irradiated fuel in the RPV Operating Mode Applicability: Mode 5..............Refueling Emergency Action Level(s): (1 or 2)
- 2. a. Unexpected loss of RCS inventory AND
- b. RPV level cannot be monitored Basis:
This IC is included because it may be a precursor of more serious conditions and is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water level below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level lowering below the RPV flange warrants declaration of an Unusual Event due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists.
The difference between IC CU1 and IC CU2 deals with the RCS conditions that exist between refueling and other operating modes. In other operating modes the RCS will be intact and RPV normal level monitoring means are available. In the refueling mode the RCS is not intact and RPV level is monitored by different means. In the refueling mode, normal means of core temperature indication and RPV level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 19 of 110 CU2 (cont.)
EAL#1 involves a drop in RCS level below the top of the RPV flange that continues for >15 minutes due to an UNPLANNED event. This EAL is not applicable to reductions in flooded reactor cavity level (covered by IC AU2 EAL1) until such time as the level drops to the level of the vessel flange. If RPV level continues to drop and reaches the Low-Low ECCS Actuation Setpoint then escalation to CA2 would be appropriate.
Continued loss of RCS inventory will result in escalation to the Alert level via either IC CA2 (Loss of RCS Inventory with Irradiated Fuel in the RPV) or CA3 (Inability to maintain plant in Cold Shutdown with irradiated fuel in the RPV).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 20 of 110 CU3 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s): (1 or 2)
This IC is included as a NOUE because it may be a precursor of more serious conditions and is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered. In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). In addition, the operators should be able to monitor RCS temperature and RPV level so that escalation to the alert level via CA3 or CA1/CA2 will occur if required.
During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions that lower water level below the RPV flange are carefully planned and procedurally controlled.
Loss of forced decay heat removal at reduced inventory may result in more rapid rises in RCS temperatures depending on the time since shutdown. Escalation to the Alert level via CA3 is provided should an UNPLANNED event result in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 20 minutes with CONTAINMENT CLOSURE not established.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 21 of 110 CU3 (cont.)
Unlike the cold shutdown mode, normal means of core temperature indication and RPV level indication may not be available in the refueling mode. Redundant means of RPV level indication are therefore procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the cold shutdown of refueling modes, EAL 2 would result in declaration of a NOUE if either temperature or level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via CA2 based on an inventory loss or CA3 based on exceeding its temperature criteria.
The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 22 of 110 CU5 Initiating Condition -- UNUSUAL EVENT Loss of all offsite power to Div I & II ESF busses for >15 Minutes Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s):
- 1. a. Loss of power to all of the following transformers for > 15 minutes:
ESF-11 ESF-21 ESF-12 AND
- b. At least Div. I or Div. II Diesel Generators are supplying power to emergency busses.
Basis:
Prolonged loss of offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., Station Blackout). This IC is met even if all emergency diesel generators start and provide AC power to the ESF busses. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the event to an Alert, due to subsequent loss of the diesel generators, will occur in accordance with IC CA5, (Loss of all offsite and onsite AC power to Div I & II ESF busses).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 23 of 110 CU6 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of required DC power for >15 Minutes Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s):
- 1. a. UNPLANNED loss of Vital DC power to required DC busses based on bus <105 VDC bus voltage indications AND
- b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.
Basis:
The purpose of this IC and its associated EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.
UNPLANNED is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a Train related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per IC CA3 (Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV).
The 105 VDC is based on the minimum bus voltage necessary for the operation of safety related equipment.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 24 of 110 CU7 Initiating Condition -- UNUSUAL EVENT Inadvertent criticality OPERATING MODE APPLICABILITY Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s):
- 1. An UNPLANNED extended positive period observed on nuclear instrumentation.
Basis:
This IC addresses inadvertent criticality events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States) such as fuel mis-loading events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated) which are addressed in the companion IC SU10.
This condition can be identified using period monitors. The term extended is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration. These short term positive periods are the result of the rise in neutron population due to subcritical multiplication.
Escalation of the event to an Alert will occur in accordance with IC HA2 (Other conditions exist which in the judgment of the Emergency Director warrants declaration of an Alert).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 25 of 110 CU8 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of all onsite or offsite communications capabilities Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s): (1 or 2)
- 1. Loss of all onsite communications capability affecting the ability to perform routine operations (See Table C2)
- 2. Loss of all offsite communications capability (See Table C3)
Table C2 Onsite Communications Equipment Plant Radio System Plant Paging System Sound Powered Phones In-plant telephones Table C3 Offsite Communications Equipment All telephone lines (commercial and fiber optic)
Satellite telephone OHL NRC telephones (ENS, HPN, MCL, RSCL, PMCL)
UHF radios Basis:
The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 26 of 110 CU8(cont)
The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g.,
relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.
The list for onsite communications loss encompasses the loss of all means of routine communications (e.g., commercial telephones, sound powered phone systems, page party system (Gaitronics) and radios /
walkie talkies).
The list for offsite communications loss encompasses the loss of all means of communications with offsite authorities. This should include the ENS, commercial telephone lines, telecopy transmissions, and dedicated phone systems.
There is no escalation above the Unusual Event for this event.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 27 of 110 CA1 Initiating Condition -- ALERT Loss of RCS inventory Operating Mode Applicability: Mode 4..............Cold Shutdown Emergency Action Level(s): (1 or 2)
- 2. a. Unexpected loss of RCS inventory AND
- b. RPV level cannot be monitored for > 15 minutes Basis:
These EALs serve as precursors to a loss of ability to adequately cool the fuel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level drop and potential core uncovery. This condition will result in a minimum classification of Alert. The Low-Low ECCS Actuation Setpoint was chosen because it is a standard setpoint at which all available injection systems automatically start. The reaching of this setpoint would therefore be indicative of a failure of the RCS barrier.
In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 28 of 110 CA1(cont)
In the cold shutdown mode, normal RPV level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring. The 15-minute duration for the loss of level indication was chosen because it is half of the IC CS1 (Loss of RCS inventory affecting core decay heat removal capability) duration. The 15-minute duration allows CA1 to be an effective precursor to CS1. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CS1 basis. Therefore this EAL meets the definition for an Alert emergency.
The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.
If RPV level continues to drop then escalation to Site Area will be via CS1 (Loss of Inventory affecting core decay heat removal capability).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 29 of 110 CA2 Initiating Condition -- ALERT Loss of RCS inventory with irradiated fuel in the RPV Operating Mode Applicability: Mode 5..............Refueling Emergency Action Level(s): (1 or 2)
- 2. a. Unexpected loss of RCS inventory AND
- b. RPV level cannot be monitored for > 15 minutes Basis:
These EALs serve as precursors to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level drop and potential core uncovery. This condition will result in a minimum classification of Alert. The Low-Low ECCS Actuation Setpoint was chosen because it is a standard setpoint at which all available injection systems automatically start. The reaching of this setpoint would therefore be indicative of a failure of the RCS barrier.
In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 30 of 110 CA2(cont)
In the refueling mode, normal means RPV level instrumentation indication may not be available.
Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring. The 15-minute duration for the loss of level indication was chosen because it is half of the IC CS2 (Loss of RCS inventory affecting core decay heat removal capability with irradiated fuel in the RPV) duration. The 15-minute duration allows CA2 to be an effective precursor to CS2. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CS2 basis. Therefore this EAL meets the definition for an Alert emergency.
The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.
If RPV level continues to drop then escalation to Site Area will be via CS2 (Loss of Inventory affecting core decay heat removal capability with irradiated fuel in the RPV).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 31 of 110 CA3 Initiating Condition -- ALERT Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s): (1 or 2 or 3)
- 1. With CONTAINMENT CLOSURE not established and RCS integrity not established an UNPLANNED event results in RCS temperature exceeding 200 °F OR
- 2. With CONTAINMENT CLOSURE established and RCS integrity not established an UNPLANNED event results in RCS temperature exceeding 200 °F for >20 minutes See Note OR
- 3. An UNPLANNED event results in RCS temperature exceeding 200 °F for >60 minutes See Note or results in RPV pressure rise > 10 psig.
Note: If a RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable Basis:
EAL 1 addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are established. RCS integrity is in place 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). No delay time is allowed for EAL 1 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 32 of 110 CA3(cont)
EAL 2 addresses the complete loss of functions required for core cooling for > 20 minutes during refueling and cold shutdown modes when CONTAINMENT CLOSURE is established but RCS integrity is not established. As in EAL 1, RCS integrity should be assumed to be in place 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). The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible. The allowed time frame is consistent with the guidance provided by Generic Letter 88-17, "Loss of Decay Heat Removal" and is believed to be conservative given that a low pressure containment barrier to fission product release is established. The Note indicates that EAL 2 is not applicable if actions are successful in restoring a RCS heat removal system to operation and RCS temperature is being reduced within the 20 minute time frame.
EAL 3 addresses complete loss of functions required for core cooling for > 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in EAL 1 and 2, RCS integrity should be considered to be in place 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). The status of CONTAINMENT CLOSURE in this EAL is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety. The >10 psig pressure rise covers situations where, due to high decay heat loads, the time provided to restore temperature control, could be less than 60 minutes. The Note indicates that EAL 3 is not applicable if actions are successful in restoring a RCS heat removal system to operation and RCS temperature is being reduced within the 60 minute time frame assuming that the RPV pressure rise has remained less than 10 psig A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary UNPLANNED excursion above 200°F when the heat removal function is available.
Escalation to a Site Area Emergency would be via CS1 or CS2 should boiling result in significant RPV level loss leading to core uncovery.
The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 33 of 110 CA5 Initiating Condition -- ALERT Loss of all offsite and Loss of all onsite AC Power to Div I & II ESF Busses Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. a. Loss of power to all of the following transformers:
ESF-11 ESF-21 ESF-12 AND
- b. Failure of both Div. I and Div. II Diesel Generators to supply power to emergency busses.
AND
- c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.
Basis:
Loss of all AC power to Div I&II compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, raising the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency IC SS1, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 34 of 110 CS1 Initiating Condition -- SITE AREA EMERGENCY Loss of RCS inventory affecting core decay heat removal capability Operating Mode Applicability: Mode 4..............Cold Shutdown Emergency Action Level(s): (1 or 2)
- 1. With CONTAINMENT CLOSURE not established:
- 2. With CONTAINMENT CLOSURE established:
- b. RPV level cannot be monitored for >30 minutes with either:
Unexpected loss of RCS inventory.
Erratic Source Range Monitor indication
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 35 of 110 CS1(cont)
Basis:
Under the conditions specified by this IC, continued reduction in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV.
In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).
In the cold shutdown mode, normal RPV level indication systems will normally be available. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring.
These EALs are based on 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. A number of variables such as initial vessel level, or shutdown heat removal system design can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30-minutes was chosen.
The 30-minute duration allowed when CONTAINMENT CLOSURE is established allows sufficient time for actions to be performed to recover needed cooling equipment and is considered to be conservative given that level is being monitored via CS1 and CS2. For BWRs effluent releases would be monitored and escalation would be via Category A ICs, if required.
Escalation to a General Emergency is via CG1 (Loss of RPV inventory affecting fuel clad integrity with containment challenged with irradiated fuel in the RPV) or radiological effluent IC AG1 (Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the actual or projected duration of the release using actual meteorology).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 36 of 110 CS2 Initiating Condition -- SITE AREA EMERGENCY Loss of RCS inventory affecting core decay heat removal capability with irradiated fuel in the RPV Operating Mode Applicability: Mode 5..............Refueling Emergency Action Level(s): (1 or 2)
- 1. With CONTAINMENT CLOSURE not established:
- b. RPV level cannot be monitored with indication of core uncovery as evidenced by:
Containment High Range Radiation Monitor reading > 100 R/hr.
Erratic Source Range Monitor indication.
- 2. With CONTAINMENT CLOSURE established:
- b. RPV level cannot be monitored with indication of core uncovery as evidenced by:
Containment High Range Radiation Monitor reading > 100 R/hr.
Erratic Source Range Monitor indication.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 37 of 110 CS2(cont)
Basis:
Under the conditions specified by this IC, continued reduction in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV. Since BWRs have RCS penetrations below the setpoint, continued level reduction may be indicative of pressure boundary leakage.
In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).
These EALs are based on 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. A number of variables such as initial vessel level, or shutdown heat removal system design can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30-minutes was chosen.
As water level in the RPV lowers, the dose rate above the core will rise. Based on the GE results in NEDC-33045P, a radiation level of 200 R/hr would be expected in the containment in the event of 1%
cladding failure with full power core source terms and one hour of decay. Therefore, even conservatively neglecting the direct shine from the reactor core, a radiation level of 100 R/hr would be expected before any significant cladding damage had occurred during Mode 5. Additionally, post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
For EAL 2 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.
For BWRs effluent releases would be monitored and escalation would be via Category A ICs, if required.
Escalation to a General Emergency is via CG1 (Loss of RPV inventory affecting fuel clad integrity with containment challenged with irradiated fuel in the RPV) or radiological effluent IC AG1 (Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the actual or projected duration of the release using actual meteorology).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 38 of 110 CG1 Initiating Condition -- GENERAL EMERGENCY Loss of RCS inventory affecting fuel clad integrity with containment challenged with irradiated fuel in the RPV Operating Mode Applicability: Mode 4..............Cold Shutdown Mode 5..............Refueling Emergency Action Level(s): (1 and 2 and 3)
- 1. Unexpected loss of RCS.
AND
- 2. RPV level:
- b. cannot be monitored with indication of core uncovery for >30 minutes as evidenced by one or more of the following:
Containment High Range Radiation Monitor reading > 100 R/hr.
Erratic Source Range Monitor indication AND
- 3. Indication of CONTAINMENT challenged as indicated by one or more of the following:
Explosive mixture inside Containment Containment pressure >15 psig CONTAINMENT CLOSURE not established Secondary Containment area radiation monitor above the EOP Max Safe Operating Value in Table C1.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 39 of 110 CG1(cont)
Table C1 Area Max Safe Value RHR Room A 8 x 104 mr/hr RHR Room B 8 x 104 mr/hr RHR HX A Hatch 8 x 104 mr/hr RHR HX B Hatch 8 x 104 mr/hr RCIC Room 8 x 104 mr/hr MSL Rad Monitor 8 x 104 mr/hr SGTS Fltr. Trn. 8 x 102 mr/hr Basis:
For EAL 1 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring.
For EAL 1 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost, the operators would need to determine that RCS inventory loss was occurring.
EAL 2 represents the inability to restore and maintain RPV level to above the top of active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level.
These EALs are based on 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. A number of variables such as initial vessel level, or shutdown heat removal system design can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30-minutes was chosen.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 40 of 110 CG1(cont)
As water level in the RPV lowers, the dose rate above the core will rise. Based on the GE results in NEDC-33045P, a radiation level of 200 R/hr would be expected in the containment in the event of 1%
cladding failure with full power core source terms and one hour of decay. Therefore, even conservatively neglecting the direct shine from the reactor core, a radiation level of 100 R/hr would be expected before any significant cladding damage had occurred during Mode 4 or 5. Additionally, post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
The General Emergency is declared on the occurrence of the loss or imminent loss of function of all three barriers. Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the CONTAINMENT breached or challenged, then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a General Emergency.
In the context of EAL 3 CONTAINMENT CLOSURE is the action taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. CONTAINMENT CLOSURE should not be confused with refueling containment integrity as defined in Technical Specifications. Shutdown contingency plans typically provide for re-establishing CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory functions. If the closure is re-established prior to exceeding the temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier EALs, escalation to General Emergency would not occur.
The pressure at which containment is considered challenged is the Containment design pressure of 15 psig.
The EP-4 Maximum Safe radiation monitor values are used are used to provide indication of elevated release that may be indicative of a challenge to secondary containment.
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 gasses in CONTAINMENT. However, CONTAINMENT monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 41 of 110 E-HU1 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3)
- 1. Natural phenomena events affecting a loaded cask CONFINEMENT BOUNDARY.
Tornado Hurricane force winds Earthquake Flood Lightning Extreme Environmental Temperature OR
- 2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.
Dropped cask Cask Tip-over Explosive Overpressure Damage by Missile generated by Natural Phenomena Fire Damage Blockage of air vents and/or burial under debris OR
- 3. Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage CONFINEMENT BOUNDARY.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 42 of 110 E-HU1(cont)
Basis:
An UNUSUAL EVENT in this IC is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on loaded fuel cask CONFINEMENT BOUNDARY loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.
The accidents analyzed in Chapter XX of the XXXXXX FSAR show that the Multi-Purpose Canister (MPC) confinement boundary withstands all credible design basis postulated accidents and natural phenomena events.
EAL #1 addresses response to a natural phenomena event affecting a cask. The loss or degradation of the CONFINEMENT BOUNDARY must be determined by a visual and radiological inspection after the event to meet this EAL. Other procedures for severe weather, operating procedures and alarm response instructions provide guidance on the identification of a natural event such as tornado or hurricane. The basis for the seismic event is the ISFSI design basis earthquake. The extreme environmental temperature is XXX° F and is assumed to exist for a sufficient duration to allow the system to achieve thermal equilibrium.
EAL #2 addresses response to a dropped cask, a tipped over cask, explosion, missile damage, or fire damage. The loss or degradation of the CONFINEMENT BOUNDARY must be determined by a visual and/or radiological inspection after the event to meet this EAL. The only concern for a fire is related to a transport vehicle fuel tank fire engulfing the loaded cask while it is being moved to the ISFSI.
EAL #3 is intended to address any condition not explicitly detailed as an EAL threshold value, which, in the judgment of the Emergency Director, is a potential degradation in the level of safety of the ISFSI.
Emergency Directors judgment is to be based on known conditions and expected response to mitigating activities within a short time period.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 43 of 110 E-HU2 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed Security Event with potential loss of safety of the ISFSI.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. Security event as determined from the GGNS Security Plan for ISFSI and reported by the GGNS Security Shift Supervision.
Basis:
The UNUSUAL EVENT is based on the GGNS Security Plan for ISFSI. Security events which do not represent a potential degradation in the level of safety of the ISFSI, are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72.
Reference is made to GGNS Security Shift Supervision because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Security Plan.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 44 of 110 FU1 Initiating Condition -- UNUSUAL EVENT ANY Loss or ANY Potential Loss of Primary Containment Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. ANY Loss or ANY Potential Loss of Primary Containment Basis:
The Primary Containment barrier includes connections up to and including the outermost containment isolation valves. Primary Containment barrier EALs are used primarily as discriminators for escalation from an Alert to a Site Area Emergency or a General Emergency.
See Primary Containment Parameters section for additional information.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 45 of 110 FA1 Initiating Condition -- ALERT ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS Basis:
The Fuel Clad barrier is the zircalloy tubes that contain the fuel pellets. The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves. The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment barrier.
See Fuel Clad or Reactor Coolant System Parameters sections for additional information.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 46 of 110 FS1 Initiating Condition - SITE AREA EMERGENCY Loss or Potential Loss of ANY two Barriers Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss or Potential Loss of ANY two Barriers Basis:
Loss of two Fission Product Barriers represents a major failure of plant systems needed to protect public health and safety.
See Primary Containment, Fuel Clad or Reactor Coolant System Parameters sections for additional information.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 47 of 110 FG1 Initiating Condition - GENERAL EMERGENCY Loss of ANY two Barriers AND Loss or Potential Loss of third barrier Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss of ANY two Barriers AND Loss or Potential Loss of third barrier Basis:
Conditions required causing loss of two Fission Product Barriers AND Loss or Potential Loss of the third barrier could reasonably be expected to cause or create a potential to cause a radiological release exceeding the EPA Protective Action Guidelines.
See Primary Containment, Fuel Clad or Reactor Coolant System Parameters sections for additional information.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 48 of 110 Fuel Clad Parameters Parameter:
FC1 - Primary Coolant Activity Level Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Coolant activity >300 µCi/gm dose equivalent I131 Potential Loss: None Basis:
LOSS:
This amount of coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.
POTENTIAL LOSS:
There is no equivalent "Potential Loss" EAL for this item.
Reference:
Calculation XC-Q1111-04006 Safety Analysis Parameters in the EALs
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 49 of 110 Fuel Clad Parameters Parameter:
FC2 - RPV Water Level Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: RPV water level < -192 in.
Potential Loss: RPV water level < -167 in.
Basis:
LOSS:
The -192 in. value corresponds to the level which is used in EOPs to indicate challenge of core cooling.
This corresponds to the Minimum Steam Cooling RPV Water Level. This is the minimum RPV water that assures maximum peak cladding temperature will not exceed 1500°F. This temperature is considered to be the cladding perforation threshold below which loss of cladding integrity due to oxidation pitting is not expected to occur.
POTENTIAL LOSS:
The -167in. value corresponds to the Top of Active Fuel RPV Water Level and is the same level value used in the Reactor Coolant System barrier "Loss" EAL. If there is indication of a leak in the drywell, this level indicates a Loss of Reactor Coolant System and a Potential Loss of the Fuel Clad barriers.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 50 of 110 Fuel Clad Parameters Parameter:
FC3 - Drywell Radiation Monitoring Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Drywell Radiation monitor reading > 3000 R/hr Potential Loss: None Basis:
LOSS:
A reading in excess of 3000 R/hr indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier loss EAL RC4. Thus, this EAL indicates loss of both Fuel Clad barrier and RCS barrier.
POTENTIAL LOSS:
There is no "Potential Loss" EAL associated with this item.
Reference:
Calculation XC-Q1111-04006 Safety Analysis Parameters in the EALs
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 51 of 110 Fuel Clad Parameters Parameter:
FC4 - Emergency Director Judgment Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Any condition in the opinion of the Emergency Director that indicates a loss of the Fuel Clad barrier Potential Loss: Any condition in the opinion of the Emergency Director that indicates a potential loss of the Fuel Clad barrier Basis:
LOSS or POTENTIAL LOSS:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. The inability to monitor the barrier should also be considered in this EAL as a factor in the Emergency Directors judgment that the barrier may be considered lost or potentially lost. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 52 of 110 Reactor Coolant System Parameters Parameter:
RC1 - Drywell Pressure Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Pressure >1.39 psig with indications of a reactor coolant leak in the drywell Potential Loss: None Basis:
LOSS:
The drywell pressure is based on the drywell high pressure set point [1.39 psig] which indicates a LOCA and automatically initiates the ECCS. The requirement to determine that a leak exists concurrently with the high pressure in the drywell assures that this EAL is limited to actual loss of Reactor Coolant System barrier and not implemented during testing, loss of drywell cooling, etc.
POTENTIAL LOSS:
There is no "Potential Loss" EAL corresponding to this item.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 53 of 110 Reactor Coolant System Parameters Parameter:
RC2 - RPV Water Level Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: RPV water level < -167 in. with indications of a reactor coolant leak in the drywell Potential Loss: None Basis:
LOSS:
The -167 in. value is the same as the Fuel Clad barrier "Potential Loss" EAL and corresponds to the Top of Active Fuel RPV Water Level. This EAL appropriately escalates the emergency class to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS barrier and a potential loss of the Fuel Clad barrier. The requirement to determine that a leak exists concurrently with the RPV low level assures that this EAL is limited to actual loss of Reactor Coolant System Barrier events making it consistent with other EALs in this category. (See Reactor Coolant System, Drywell Pressure EAL).
POTENTIAL LOSS:
There is no "Potential Loss" EAL corresponding to this item.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 54 of 110 Reactor Coolant System Parameters Parameter:
RC3 - Reactor Coolant System Leak Rate Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Unisolable MSL break as indicated by the failure of both MSIVs in any one line to close AND High MSL Flow and High Steam Tunnel Temperature annunciators OR Direct report of steam release Potential Loss: Reactor Coolant System leakage > 50 gpm inside the drywell.
OR Unisolable primary system leakage outside Primary Containment as indicated by any Area Temperature or Area Radiation > Alert Value in Table F1.
Basis:
LOSS:
An unisolable Main Steam Line break is a breach of the RCS barrier. Thus, this EAL is included for consistency with the Alert emergency classification.
POTENTIAL LOSS:
The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak, however, break propagation leading to significantly larger loss of inventory is possible. If primary system leak rate information is unavailable, other indicators of RCS leakage should be used.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 55 of 110 Reactor Coolant System Parameters RC3 - Reactor Coolant System Leak Rate (cont)
Potential loss of Reactor Coolant System based on primary system leakage outside the primary containment is determined from Table F1. The Alert Limit column contains area temperatures and area radiation values in the areas of the main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment. These values are the same as the Operating Limit in Table 3 of EP-4. The indicators should be confirmed to be caused by Reactor Coolant System leakage in the area.
The area temperature or radiation values in the Alert Limit column are to be used to declare an Alert classification if no other barrier degradation is observed.
The GE / SAE Limit column contains the same values as the Max Safe Value in Table 3 of EP-4. The values in this column should be used when escalating to Site Area Emergency or General Emergency as multiple barriers degrade.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 56 of 110 Reactor Coolant System Parameters Parameter:
RC4 - Drywell Radiation Monitoring Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Drywell Radiation monitor reading > 100 R/hr with indications of a leak in the drywell.
Potential Loss: None Basis:
LOSS:
The Drywell Radiation monitor reading indicates the release of reactor coolant to the drywell. This reading will be less than that specified for Fuel Clad barrier EAL FC3. This arbitrary value was chosen to be well in excess of expected radiation readings during normal plant operations. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading elevated to that value specified by FC3, fuel damage would also be indicated.
POTENTIAL LOSS:
There is no "Potential Loss" EAL corresponding to this item.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 57 of 110 Reactor Coolant System Parameters Parameter:
RC5 - Emergency Director Judgment Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Any condition in the opinion of the Emergency Director that indicates a loss of the RCS barrier.
Potential Loss: Any condition in the opinion of the Emergency Director that indicates a potential loss of the RCS barrier.
Basis:
LOSS or POTENTIAL LOSS:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. The inability to monitor the barrier should also be considered in this EAL as a factor in the Emergency Directors judgment that the barrier may be considered lost or potentially lost. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 58 of 110 Primary Containment Parameters Parameter:
PC1 - Primary Containment Pressure Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Rapid unexplained loss of pressure following initial pressure rise.
OR Pressure response not consistent with LOCA conditions.
Potential Loss: >15 psig and rising in Primary Containment.
OR Ctmt H2 concentration in HDOL Unsafe Zone.
Basis:
LOSS:
Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity. Containment pressure should rise as a result of mass and energy release into containment from a LOCA. Thus, containment pressure not rising under these conditions indicates a loss of containment integrity. This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated. The unexpected response is important because it is the indicator for a containment bypass condition. Due to the size of the Mark III containment small breaks may not be able to be detected by this EAL. Smaller breaks are addressed in the Primary Ctmt Isolation Failure or Bypass EAL POTENTIAL LOSS:
The 15 psig value used for potential loss of containment is based on the containment design pressure.
Existence of an explosive mixture means a hydrogen and oxygen concentration of at least the lower deflagration limit exists.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 59 of 110 Primary Containment Parameters Parameter:
PC2 - RPV Water Level Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: None Potential Loss: Primary Containment flooding required by:
SAP 1, 3, 4, 5 or 6.
Basis:
LOSS:
There is no "Loss" EAL corresponding to this item.
POTENTIAL LOSS:
The entry into the Severe Accident Procedures (SAPs) indicates reactor vessel water level can not be restored and that a core melt sequence is in progress. EOPs direct the operators to enter SAPs when Reactor Vessel Level cannot be restored and maintained greater than -192 in. or is unknown. Entry into the SAPs is a logical escalation in response to the inability to maintain reactor vessel level.
The conditions in this potential loss EAL represent imminent core melt sequences which, if not corrected, could lead to vessel failure and elevated potential for containment failure. In conjunction with and an escalation of the level EALs in the Fuel Clad and RCS barrier columns, this EAL will result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the EOPs have been ineffective in restoring reactor vessel level above the RCS and Fuel Clad barrier threshold values, there is not a success path and a core melt sequence is in progress. Entry into the SAPs is a logical escalation in response to the inability to maintain reactor vessel level.
Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow SAPs to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 60 of 110 Primary Containment Parameters Parameter:
PC3 - Primary Containment Isolation Failure or Bypass Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Failure to isolate any penetration AND downstream pathway to the environment exists.
OR Intentional venting required by EPs or SAPs.
OR Unisolable primary system leak outside Primary Containment as indicated by any Area Temperature or Area Radiation level >SAE/GE Value in Table F-1.
Potential Loss: None Basis:
LOSS:
This EAL is intended to cover the inability to isolate the containment when containment isolation is required. The qualifier AND downstream pathway to the environment exists is included so as to properly define the particular situation (i.e., system breach or direct path). If no downstream pathway to the environment exists, then Primary Containment is not lost. In addition, the presence of area radiation or temperature alarms indicating unisolable primary system leakage outside the drywell are covered after a containment isolation. The indicators should be confirmed to be caused by RCS leakage.
An intentional venting of primary containment for pressure control per EPs or SAPs to the secondary containment and/or the environment is considered a loss of containment due to the potential for a large radioactivity inventory to be present in the containment. Containment venting for temperature or pressure control when not in an accident situation should not be considered.
POTENTIAL LOSS:
There is no "Potential Loss" EAL corresponding to this item.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 61 of 110 Primary Containment Parameters Parameter:
PC4 - Significant Radioactive Inventory in Primary Containment Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: None Potential Loss: Containment Radiation monitor reading > 10,000 R/Hr Basis:
LOSS:
There is no "Loss" EAL corresponding to this item.
POTENTIAL LOSS:
This reading is a value which indicates significant fuel damage well in excess of that required for loss of RCS or Fuel Clad. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in the primary containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates that such conditions do not exist when the amount of clad damage is less than 20%. Therefore, the reading corresponds to 20% fuel clad damage.
Reference:
Calculation XC-Q1111-04006 Safety Analysis Parameters in the EALs
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 62 of 110 Primary Containment Parameters Parameter:
PC5 - Emergency Director Judgment Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
Loss: Any condition in the opinion of the Emergency Director that indicates a loss of the Primary Containment barrier.
Potential Loss: Any condition in the opinion of the Emergency Director that indicates a potential loss of the Primary Containment barrier.
Basis:
LOSS or POTENTIAL LOSS:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the Primary Containment barrier is lost or potentially lost. The inability to monitor the barrier should also be considered in this EAL as a factor in the Emergency Directors judgment that the barrier may be considered lost or potentially lost. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 63 of 110 HU1 Initiating Condition - UNUSUAL EVENT Confirmed security event which indicates a potential degradation in the level of safety of the plant Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. Security events as determined from the GGNS Safeguards Contingency Plan and reported by the GGNS security shift supervision.
- 2. A credible site specific security threat notification.
Basis:
Reference is made to GGNS security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred.
Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Safeguards Contingency Plan.
EAL 1 is based on GGNS Site Security Plans. Security events which do not represent a potential degradation in the level of safety of the plant, are reported under 10CFR73.71 or in some cases under 10CFR50.72. Examples of security events that indicate Potential Degradation in the Level of Safety of the Plant are provided below for consideration.
Consideration should be given to the following types of events when evaluating an event against the criteria of the site specific Security Contingency Plan: SABOTAGE, HOSTAGE/EXTORTION, CIVIL DISTURBANCE, and STRIKE ACTION.
The intent of EAL 2 is to ensure that appropriate notifications for the security threat are made in a timely manner. Only the plant to which the specific threat is made need declare an UNSUAL EVENT.
The determination of credible is made by the Shift Manager (Emergency Director) with assistance from the Security Superintendent. Additional resources may be necessary to assist in this determination.
These additional resources could include other sites, corporate, law enforcement or federal resources.
The threat must be specific to Grand Gulf.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 64 of 110 HU1(cont)
Site Security Code Yellow (Armed adversary attempting to or has entered company property) is limited to security events involving an armed attack against the plant. An armed attack against the plant is a unique security emergency that is expected to be an extremely fast moving event which presents an immediate and serious threat to human life. This site security code system is used to enhance communication for this event and allows immediate recognition and rapid notification for this event. It is imperative for personnel to take cover immediately to minimize loss of life. A Site Security Code Orange would result in escalation to an ALERT.
INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE would result in EAL escalation to an ALERT.
A higher initial classification could be made based upon the nature and timing of the threat and potential consequences. The licensee shall consider upgrading the emergency response status and emergency classification in accordance with the GGNS Safeguards Contingency Plan and Emergency Plans.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 65 of 110 HU2 Initiating Condition - UNUSUAL EVENT Other Conditions existing which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occur.
Basis:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the plant conditions warrant declaration of an Unusual Event.
The inability to monitor identified parameters should also be considered in this EAL as a factor in the Emergency Directors judgment that the hazard actually exists. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 66 of 110 HU4 Initiating Condition -- UNUSUAL EVENT FIRE within PROTECTED AREA boundary not extinguished within 15 minutes of detection.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. FIRE in buildings or areas contiguous to any VITAL AREA (Table H3) not extinguished within 15 minutes of control room notification or verification of a control room alarm Table H3 Buildings or Areas Contiguous To Any VITAL AREAS Unit I Containment Unit I & II Auxiliary Building Control Building Unit I & II Turbine Building Diesel Generator Rooms SSW Pump & Valve Rooms
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 67 of 110 HU4(cont)
Basis:
The purpose of this IC is to address the magnitude and extent of FIREs that may be potentially significant precursors to damage to safety systems. As used here, Detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a FIRE is occurring, or indication of a VALID fire detection system alarm. Verification of a fire detection system alarm includes actions that can be taken within the control room or other nearby locations to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.
The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREs that are readily extinguished (e.g., smoldering waste paper basket). The intent of this IC is not to include buildings (i.e., warehouses) or areas that are not contiguous (in actual contact with or immediately adjacent) to plant Vital Areas. This excludes FIREs within administration buildings, waste-basket FIREs, and other small FIREs of no safety consequence.
Escalation to a higher emergency class is by IC HA4, "FIRE affecting the operability of plant safety systems required to establish or maintain safe shutdown".
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 68 of 110 HU5 Initiating Condition - UNUSUAL EVENT Release of toxic or flammable gases deemed detrimental to normal operation of the plant.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect NORMAL PLANT OPERATIONS.
- 2. Report by Local, County/Parish or State Officials for evacuation or sheltering of site personnel based on an offsite event.
Basis:
This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter the site boundary and affect normal plant operations. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. The EALs are intended to not require significant assessment or quantification.
The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety.
Escalation of this EAL is via HA5, which involves a quantified release of toxic or flammable gases affecting VITAL AREAs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 69 of 110 HU6 Initiating Condition -- UNUSUAL EVENT Natural and destructive phenomena affecting the PROTECTED AREA Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3 or 4 or 5 or 6 or 7)
- 1. Valid indication of a felt earthquake:
Vibratory ground motion felt in the PROTECTED AREA and recognized as an earthquake.
AND Activated seismic switches as indicated by activation of the Seismic Monitoring System:
Strong Motion Accelerometer System Activation (P856-1A-A1)
- 2. Report by plant personnel of tornado striking within PROTECTED AREA boundary.
- 3. Vehicle crash into Plant Structures containing Functions or Systems Required for Safe Shutdown Table H2 within PROTECTED AREA boundary.
- 4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to a permanent structure or equipment.
- 5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 6. Uncontrolled flooding in the Auxiliary Building (Table H1) that has the potential to affect safety related equipment needed for the current operating mode.
- 7. Severe weather with indication of sustained high winds 74 mph within PROTECTED AREA boundary.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 70 of 110 HU6(cont)
Table H1 Auxiliary Building Area Parameters Area Max Safe Operating Value RHR Room A 93 FT. 6 IN. (P870-2A-E1)
RHR Room B 93 FT. 6 IN. (P870-10A-G1)
RHR Room C 93 FT. 6 IN. (P870-10A-G2)
RCIC Room 93 FT. 6 IN. (P870-2A-A1)
LPCS Room 93 FT. 6 IN. (P870-2A-F1)
HPCS Room 93 FT. 6 IN. (P870-5A-H1)
Table H2 Structures Containing Functions or Systems Required for Safe Shutdown Unit I Containment Unit I Auxiliary Building Control Building Unit 1 Turbine Building Diesel Generator Rooms SSW Pump & Valve Rooms Basis:
The EALs in this IC are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Areas identified in the EALs define the location of the event based on the potential for damage to equipment contained therein. Escalation of the event to an Alert occurs when the magnitude of the event is sufficient to result in damage to equipment contained in the specified location.
EAL #1 damage may be caused to some portions of the site, but should not affect ability of safety functions to operate. Method of detection can be based on in instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:
An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 71 of 110 HU6(cont)
EAL #2 is based on the assumption that a tornado striking (touching down) within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.
EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to have caused damage in a plant VITAL AREA, the event may be escalated to Alert.
EAL #4 only those EXPLOSIONS of sufficient force to damage permanent structures or equipment within the PROTECTED AREA should be considered. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the EXPLOSION, if applicable.
EAL #5 is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual FIREs and flammable gas build up are appropriately classified via HU4 and HU5.
Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant. This EAL is consistent with the definition of an UNUSUAL EVENT while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or by the radiological releases. These latter events would be classified by the Radiological ICs or Fission Product Barrier ICs.
EAL #6 addresses the effect of flooding caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant, which are not designed to be wetted or submerged.
Escalation of the emergency classification is based on the damage caused or by access restrictions that prevent necessary plant operations or systems monitoring.
EAL #7 is based on the assumption that high winds within the PROTECTED AREA may have potentially damaged plant structures, listed in Table H2, containing functions or systems required for safe shutdown of the plant. The high wind site specific value is based on the wind speed (74 mph) to classify severe weather conditions as a hurricane. FSAR design basis is that all Seismic Category I structures at GGNS are designed to withstand 90 mph fastest mile of sustained wind from 0 to 50 ft above ground, based upon a 100-yr period of recurrence. Methods to measure wind speed in the PROTECTED AREA are not available, therefore a sustained indication of 74 mph on the Meteorological Tower 10 meter wind speed indication will be used to determine that this EAL is met. The upper scale for the 10 meter wind speed on the MET Tower is 125 mph. If the MET Tower is not operable, other sources may be considered for estimated wind speed at GGNS. If damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 72 of 110 HA1 Initiating Condition - ALERT Confirmed security event in plant PROTECTED AREA Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE OR
- 2. Security events as determined from the GGNS Safeguards Contingency Plan and reported by the GGNS security shift supervision.
Basis:
This class of security events represents an escalated threat to plant safety above that contained in the UNUSUAL EVENT. A confirmed INTRUSION report is satisfied if physical evidence indicates the presence of a HOSTILE FORCE within the PROTECTED AREA.
Consideration should be given to the following types of events when evaluating an event against the criteria of the site specific Security Contingency Plan: SABOTAGE, HOSTAGE/EXTORTION, and STRIKE ACTION. The Safeguards Contingency Plan identifies numerous events/conditions that constitute a threat/compromise to a Stations security. Only those events that involve Actual or Potential Substantial degradation to the level of safety of the plant need to be considered. The following events would not normally meet this requirement; (e.g., Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions.
Site Security Code Orange (Armed adversary attempting to or has crossed the PROTECTED AREA fence) is limited to security events involving an armed attack against the plant. An armed attack against the plant is a unique security emergency that is expected to be an extremely fast moving event which presents an immediate and serious threat to human life. A site security code system is used to enhance communication for this event and allows immediate recognition and rapid notification for this event. It is imperative for personnel to take cover immediately to minimize loss of life. A Site Security Code Red would result in escalation to a Site Area Emergency.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 73 of 110 HA1(cont)
INTRUSION into a VITAL AREA by a HOSTILE FORCE will escalate this event to a Site Area Emergency Reference is made to GGNS security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred.
Training on security event classification is closely controlled due to the strict secrecy controls placed on the plant Security Plan.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 74 of 110 HA2 Initiating Condition -- ALERT Other Conditions existing which in the judgment of the Emergency Director warrant declaration of an ALERT.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels Basis:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the plant conditions warrant declaration of an Alert. The inability to monitor identified parameters should also be considered in this EAL as a factor in the Emergency Directors judgment that the hazard actually exists. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 75 of 110 HA3 Initiating Condition -- ALERT Control room evacuation has been initiated.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. Entry into 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, for Control Room evacuation.
Basis:
With the control room evacuated, additional support, monitoring and direction through the Technical Support Center or other emergency response facilities is desirable. Inability to establish plant control from outside the control room will escalate this event to a Site Area Emergency per IC HS3.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 76 of 110 HA4 Initiating Condition -- ALERT FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. FIRE or EXPLOSION in Structures containing Functions or Systems Required for Safe Shutdown (Table H2) the following areas:
Table H2 Structures Containing Functions or Systems Required for Safe Shutdown Unit I Containment Unit I Auxiliary Building Control Building Unit 1 Turbine Building Diesel Generator Rooms SSW Pump & Valve Rooms AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified areas.
Basis:
This EAL addresses a FIRE / EXPLOSION and not the degradation in performance of affected systems.
System degradation is addressed in the System Malfunction EALs. The reference to damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate against minor FIREs /
EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs / EXPLOSIONs in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE / EXPLOSION was large enough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE /
EXPLOSION is large enough to affect more than one component.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 77 of 110 HA4(cont)
This situation is not the same as removing equipment for maintenance that is covered by a plants Technical Specifications. Removal of equipment for maintenance is a planned activity controlled in accordance with procedures and, as such, does not constitute a substantial degradation in the level of safety of the plant. A FIRE / EXPLOSION is an UNPLANNED activity and, as such, does constitute a substantial degradation in the level of safety of the plant. In this situation, an Alert classification is warranted.
The inclusion of a "report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the FIRE / EXPLOSION with reports of evidence of damage is sufficient for declaration. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform these damage assessments.
The Emergency Director also needs to consider any security aspects of the EXPLOSIONs, if applicable.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 78 of 110 HA5 Initiating Condition -- ALERT Release of toxic or flammable gases within or contiguous to a VITAL AREA which jeopardizes operation of systems required to maintain safe operations or establish or maintain safe shutdown.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. Report or detection of toxic or flammable gases in Buildings or areas Contiguous to any VITAL AREA (Table H3) in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).
- 2. Report or detection of gases in concentrations greater than the LOWER FLAMMABILITY LIMIT within Buildings or areas Contiguous to any VITAL AREA (Table H3)
Table H3 Buildings or Areas Contiguous To Any VITAL AREAS Unit I Containment Unit I & II Auxiliary Building Control Building Unit I & II Turbine Building Diesel Generator Rooms SSW Pump & Valve Rooms
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 79 of 110 HA5(cont)
Basis:
This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas contiguous to plant VITAL AREAs or other significant buildings or areas (i.e., Standby Service Water pump house). The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to plant VITAL AREAs. It is appropriate that elevated monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radioactive Effluent, or Emergency Director Judgment ICs.
EAL #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH within a VITAL AREA or any area or building contiguous to VITAL AREA. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.
EAL #2 is met when the flammable gas concentration in a VITAL AREA or any building or area contiguous to a VITAL AREA exceed the LOWER FLAMMABILITY LIMIT. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Once it has been determined that an uncontrolled release is occurring, then sampling must be done to determine if the concentration of the released gas is within this range.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 80 of 110 HA6 Initiating Condition -- ALERT Natural and destructive phenomena affecting the plant VITAL AREA Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2 or 3 or 4 or 5 or 6)
- 1. Valid indication of a seismic event greater than Operating Basis Earthquake (OBE):
Receipt of all of the following indications on SH13P856:
Containment Operating Basis Earthquake (P856-1A-A3)
Drywell Operating Basis Earthquake (P856-1A-A5)
Containment Safe Shutdown Earthquake (P856-1A-A2)
Drywell Safe Shutdown Earthquake (P856-1A-A4)
Strong Motion Accelerometer System Activation (P856-1A-A1)
AND Start of all five SMA cassette tape recorders AND Event indicator flag changes from black to white AND Event alarm yellow light in ON OR
- 2. Tornado striking within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2) or Control Room indication of degraded performance of those systems.
- 3. Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2) or Control Room indication of degraded performance of those systems.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 81 of 110 HA6(cont)
- 4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of the Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2).
- 5. Uncontrolled flooding in the Auxiliary Building (Table H1) that results in degraded safety system performance as indicated in the Control Room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.
- 6. Severe weather with indication of sustained high winds 74 mph within PROTECTED AREA boundary and resulting in VISUAL DAMAGE to Plant Structures containing Functions or Systems Required for Safe Shutdown (Table H2) or has caused damage as evidenced by Control Room indication of degraded performance of those systems.
Table H1 Auxiliary Building Area Parameters Area Max Safe Operating Value RHR Room A 93 FT. 6 IN. (P870-2A-E1)
RHR Room B 93 FT. 6 IN. (P870-10A-G1)
RHR Room C 93 FT. 6 IN. (P870-10A-G2)
RCIC Room 93 FT. 6 IN. (P870-2A-A1)
LPCS Room 93 FT. 6 IN. (P870-2A-F1)
HPCS Room 93 FT. 6 IN. (P870-5A-H1)
Table H2 Structures Containing Functions or Systems Required for Safe Shutdown Unit I Containment Unit I Auxiliary Building Control Building Unit 1 Turbine Building Diesel Generator Rooms SSW Pump & Valve Rooms
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 82 of 110 HA6(cont)
Basis:
The EALs in this IC escalate from the UNUSUAL EVENT EALs in HU6 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications may occur on the basis of other ICs (e.g., System Malfunction).
EAL #1 is based on UFSAR design basis. Seismic events of this magnitude can result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. See EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, for information on seismic event categories.
EAL #2 is based on observation of VISIBLE DAMAGE within a VITAL AREA. Wind loads of this magnitude can cause damage to safety functions.
EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant.
EAL#4 is intended to address the threat to safety related equipment imposed by missiles generated by main turbine rotating component failures. This EAL is, therefore, consistent with the definition of an Alert in that if missiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant.
EAL#5 addresses the effect of internal flooding that has resulted in degraded performance of systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety of the plant. This flooding may have been caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The areas include those areas that contain systems required for safe shutdown of the plant, that are not designed to be wetted or submerged.
EAL#6 addresses the effect of high winds within the PROTECTED AREA that has damaged plant structures, listed in Table H2, containing functions or systems required for safe shutdown of the plant.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 83 of 110 HS1 Initiating Condition - SITE AREA EMERGENCY Confirmed security event in a plant VITAL AREA Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s): (1 or 2)
- 1. INTRUSION into the plant VITAL AREA by a HOSTILE FORCE.
- 2. Security events as determined from the GGNS Safeguards Contingency Plan and reported by the GGNS security shift supervision.
Basis:
This class of security events represents an escalated threat to plant safety above that contained in the Alert IC in that a HOSTILE FORCE has progressed from the PROTECTED AREA to a VITAL AREA.
Consideration should be given to the following types of events when evaluating an event against the criteria of the site specific Security Contingency Plan: SABOTAGE and HOSTAGE / EXTORTION. The Safeguards Contingency Plan identifies numerous events/conditions that constitute a threat/compromise to a Stations security. Only those events that involve Actual of Likely Major failures of plant functions needed for protection of the public need to be considered. The following events would not normally meet this requirement; (e.g., Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions).
Site Security Code Red (Armed adversary has entered any power block building or vital area) is limited to security events involving an armed attack against the plant. An armed attack against the plant is a unique security emergency that is expected to be an extremely fast moving event which presents an immediate and serious threat to human life. A site security code system is used to enhance communication for this event and allows immediate recognition and rapid notification for this event. It is imperative for personnel to take cover immediately to minimize loss of life.
Loss of Plant Control would escalate this event to a GENERAL EMERGENCY.
Reference is made to GGNS security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred.
Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Security Plan.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 84 of 110 HS2 Initiating Condition - SITE AREA EMERGENCY Other Conditions existing which in the judgment of the Emergency Director warrant declaration of SITE AREA EMERGENCY.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. 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. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the plant conditions warrant declaration of a Site Area Emergency. The inability to monitor identified parameters should also be considered in this EAL as a factor in the Emergency Directors judgment that the hazard actually exists. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 85 of 110 HS3 Initiating Condition - SITE AREA EMERGENCY Control Room evacuation has been initiated and plant control cannot be established.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. Control Room evacuation has been initiated.
AND Control of the plant cannot be established per 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, within 15 minutes.
Basis:
Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. The intent of this IC is to capture those events where control of the plant cannot be re-established in a timely manner. The 15 minute time for transfer is based on analysis as to how quickly control must be re-established without core damage. The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director (ED) judgment. The ED is expected to make a reasonable, informed judgment within 15 minutes that the licensee has control of the plant from the remote shutdown panel.
The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink).
Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 86 of 110 HG1 Initiating Condition - GENERAL EMERGENCY Security event resulting in loss of physical control of the facility Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.
Basis:
This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of VITAL AREAs (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. Typically, these safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink). Loss of both SSW systems does not by itself constitute loss of decay heat removal capability. Example: In an extended Station Blackout condition using RCIC to maintain RPV water level above Minimum Steam Cooling Water Level, decay heat is removed by steam through an SRV to the suppression pool. If containment pressure rises due to suppression pool heat up, the containment can be vented to the environs if necessary to maintain containment pressure within EOP limits with no significant radiological consequences. If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.
This EAL should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly off-loaded reactor core in pool).
Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 87 of 110 HG2 Initiating Condition - GENERAL EMERGENCY Other conditions existing which in the judgment of the Emergency Director warrant declaration of General Emergency.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Mode 4..............Cold Shutdown Mode 5..............Refueling Mode D .............Defueled Emergency Action Level(s):
- 1. 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. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
This EAL is not intended to be used to anticipate conditions already identified in other EALs. This EAL is intended to address unanticipated conditions, not addressed explicitly elsewhere, that can be used by the Emergency Director in determining whether the plant conditions warrant declaration of a General Emergency. The inability to monitor identified parameters should also be considered in this EAL as a factor in the Emergency Directors judgment that the hazard actually exists. The emphasis is on the need for an accurate assessment recognizing that over-classification, as well as under-classification, is to be avoided.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 88 of 110 SU1 Initiating Condition -- UNUSUAL EVENT Loss of all offsite power to Div I & II ESF busses for >15 Minutes Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss of power to all of the following transformers for > 15 minutes:
ESF-11 ESF-21 ESF-12 AND At least Div. I and Div. II Diesel Generators are supplying power to emergency busses.
Basis:
Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., Station Blackout).
This IC is met even if all emergency diesel generators start and provide AC power to the ESF busses.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the event to an Alert, due to subsequent loss of diesel generators such that only one source remains, will occur in accordance with IC SA1, (AC power capability to Div I & II ESF busses reduced to a single power source for >15 minutes such that any additional single failure would result in Station Blackout).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 89 of 110 SU6 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of most or all safety system annunciation or indication in the control room for >15 minutes Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. UNPLANNED loss of most or all Control Room annunciators or indicators associated with safety systems for > 15 minutes.
Basis:
This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.
Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).
Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an elevated risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.
It is further recognized that plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure or a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the UNUSUAL EVENT is based on SU11 (Inability to reach required shutdown within Technical Specification limits).
Annunciators or indicators for this EAL include those identified the ONEPs (Off-Normal Event Procedures), and the Emergency Operating Procedures (EPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.)
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 90 of 110 SU6(cont)
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.
This UNUSUAL EVENT will be escalated to an ALERT if a SIGNIFICANT TRANSIENT is in progress during the loss of annunciation or indication.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 91 of 110 SU7 Initiating Condition -- UNUSUAL EVENT RCS leakage Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s): (1 or 2)
- 1. Unidentified or pressure boundary leakage >10 gpm.
- 2. Identified leakage >35 gpm.
Basis:
This IC is included as a UNUSUAL EVENT because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected because it is greater than the minimum detectable amount used as the Technical Specification limit and is expected to be observable with normal control room indications without lengthy calculations.
The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
Escalation of the event to an Alert will occur via Fission Product Barrier degradation ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 92 of 110 SU8 Initiating Condition -- UNUSUAL EVENT UNPLANNED loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s): (1 or 2)
- 1. Loss of all onsite communications capability affecting the ability to perform routine operations.
(See Table S1)
- 2. Loss of all offsite communications capability. (See Table S2)
Table S1 Onsite Communications Equipment Plant Radio System Plant Paging System Sound Powered Phones In-plant telephones Table S2 Offsite Communications Equipment All telephone lines (commercial and fiber optic)
Satellite telephone OHL NRC telephones (ENS, HPN, MCL, RSCL, PMCL)
UHF radios Basis:
The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10CFR50.72.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 93 of 110 SU8(cont)
The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g.,
relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.
The list for onsite communications loss encompasses the loss of all means of routine communications (e.g., commercial telephones, sound powered phone systems, page party system (Gaitronics) and radios /
walkie talkies).
The list for offsite communications loss encompasses the loss of all means of communications with offsite authorities. This should include the ENS, commercial telephone lines, telecopy transmissions, and dedicated phone systems.
There is no escalation above the Unusual Event for this event.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 94 of 110 SU9 Initiating Condition -- UNUSUAL EVENT Fuel Clad Degradation.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Offgas isolation (1H13-P601-19A-C8) due to valid Offgas Post Treatment monitor signal indicating fuel clad degradation.
- 2. Reactor coolant sample activity >4.0 µCi/gm dose equivalent I131, indicating fuel clad degradation.
Basis:
This IC is included as an UNUSUAL EVENT because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. EAL #1 addresses radiation monitor readings that provide indication of fuel clad integrity. EAL #2 addresses coolant samples exceeding coolant technical specifications for iodine spike.
Escalation of the event to an ALERT is via the Fission Product Barrier degradation monitoring ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 95 of 110 SU10 Initiating Condition -- UNUSUAL EVENT Inadvertent criticality OPERATING MODE APPLICABILITY Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. An UNPLANNED extended positive period observed on nuclear instrumentation.
Basis:
This IC addresses inadvertent criticality events. While the primary concern of this IC is criticality events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States), the IC is applicable in other modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of the plant, warranting an UNUSUAL EVENT classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated).
The Cold Shutdown/Refueling IC is CU7.
This condition can be identified using period monitors. The term extended is used in order to allow exclusion of expected short term positive periods from planned control rod movements. (Example: Short term positive periods as the result of the rise in neutron population due to subcritical multiplication)
Escalation would be by the Fission Product Barrier Matrix, as appropriate to the operating mode at the time of the event, or by Emergency Director Judgment.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 96 of 110 SU11 Initiating Condition -- UNUSUAL EVENT Inability to reach required shutdown within technical specification time limits Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time Basis:
Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a one hour report under 10CFR50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications.
An immediate UNUSUAL EVENT declaration is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an UNUSUAL EVENT is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunction, Hazards, or Fission Product Barrier degradation ICs.
There is no escalation above the UNUSUAL EVENT for this event.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 97 of 110 SA1 Initiating Condition -- ALERT AC power capability to Div I or II ESF busses reduced to a single power source for
> 15 minutes such that any additional single failure would result in Station Blackout.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. AC power capability to 15AA or 16AB reduced to a single power source for >15 minutes AND Any additional single failure will result in Station Blackout.
Basis:
This IC and the associated EALs are intended to provide an escalation from IC SU1, (Loss of All Offsite Power To Div I & II ESF Busses for > 15 Minutes). The condition indicated by this IC is the degradation of the offsite and onsite power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of one emergency generator to supply power to its emergency bus. Another related condition could be the loss of onsite emergency diesels with only one train of emergency busses being fed from offsite power.
The subsequent loss of this single power source could escalate the event to a Site Area Emergency in accordance with IC SS1, (Loss of all offsite and onsite AC power to Div I, II & III ESF busses).
Div. III Diesel Generator and Bus 17AC are not discussed explicitly in this IC. The loss of Div I & II ESF Busses are by definition a Station Blackout. If Div III Diesel Generator or 17AC is available, entry into this IC (SA1) is appropriate.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 98 of 110 SA3 Initiating Condition -- ALERT Failure of reactor protection system instrumentation to complete or initiate an automatic reactor scram once a reactor protection system setpoint has been exceeded and manual scram was successful Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Emergency Action Level(s):
- 1. Indication(s) exist that indicate that Reactor Protection System (RPS) setpoint was exceeded and RPS automatic scram was unsuccessful as evidenced by:
3/4 More than one control rod at position 02 or greater, OR 3/4 Control rod position is unknown for more than one control rod and SRMs are either upscale or count rate is rising.
AND Manual scram or ARI reduces reactor power to < 4%.
Basis:
This condition indicates failure of the automatic protection system to scram the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded.
An Alert is indicated because conditions exist that could lead to potential loss of fuel clad or RCS.
Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the automatic protection system is the issue.
A manual scram is any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (e.g., Scram Push Buttons, Reactor Mode Switch, Alternate Rod Insertion).
An alternate means to insert control rods would be required to bring the reactor sub-critical (e.g., manual reactor scram, ARI). Failure of this alternate means and failure to reduce reactor power to < 4% would escalate the event to a SITE AREA EMERGENCY. Escalation of the event to a SITE AREA EMERGENCY will occur in accordance with IC SS3, (Failure of reactor protection system instrumentation to complete or initiate an automatic reactor scram once a reactor protection system setpoint has been exceeded and manual scram was not successful).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 99 of 110 SA6 Initiating Condition -- ALERT UNPLANNED loss of most or all safety system annunciation or indication in control room with either (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory non-alarming indicators are unavailable Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. UNPLANNED loss of most or all Control Room annunciators or indicators associated with safety systems for > 15 minutes AND Either of the following: (a or b)
- a. A SIGNIFICANT TRANSIENT is in progress.
- b. Compensatory non-alarming indications are unavailable.
Basis:
This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).
"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.
Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an elevated risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide elevated monitoring of system operation.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 100 of 110 SA6(cont)
It is further recognized that plant design provides redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, an UNUSUAL EVENT should be declared based on IC SU11 (Inability to reach required shutdown within technical specification time limits).
Annunciators or indicators for this EAL include those identified the ONEPs (Off-Normal Event Procedures), and the Emergency Operating Procedures (EPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.
"Compensatory non-alarming indications" in this context includes computer based information such as SPDS. This should include any computer systems available for this use. If both a major portion of the annunciation system and all computer monitoring are unavailable, the ALERT is required.
Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.
Escalation of the event to a Site Area Emergency will occur in accordance with IC SS6 (Inability to monitor a significant transient in progress).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 101 of 110 SS1 Initiating Condition -- SITE AREA EMERGENCY Loss of all offsite and loss of all onsite AC power to Div I, II & III ESF busses Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss of power to all of the following transformers:
ESF-11 ESF-21 ESF-12 AND Failure of Div. I, II and III Diesel Generators to supply power to emergency busses.
AND Failure to restore power to to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power Basis:
Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. The 15 minute time duration was selected to exclude transient or momentary power losses.
Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus is the only energized bus then a Site Area Emergency per SS1 should be declared.
Escalation to General Emergency is via Fission Product Barrier Degradation or IC SG1, (Prolonged loss of all offsite and prolonged loss of all onsite AC power to Div I, II & III ESF busses).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 102 of 110 SS3 Initiating Condition -- SITE AREA EMERGENCY Failure of reactor protection system instrumentation to complete or initiate an automatic reactor scram once a reactor protection system setpoint has been exceeded and manual scram was NOT successful.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Emergency Action Level(s):
- 1. Indication(s) exist that indicate that Reactor Protection System (RPS) setpoint was exceeded and RPS automatic scram and a manual scram or ARI fails to reduce reactor power to < 4%.
Basis:
Automatic and manual scrams are not considered successful if action away from the reactor control console was required to scram the reactor.
Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.
A manual scram is any set of actions by the reactor operator(s) at the reactor control console which causes sufficient control rods to be rapidly inserted into the core and brings the reactor subcritical (e.g.,
Scram Push Buttons, Reactor Mode Switch, Alternate Rod Insertion).
Escalation of the event to a General Emergency will occur in accordance with IC SG2 (Failure of the reactor protection system to complete an automatic scram and manual scram was not successful and there is indication of an extreme challenge to the ability to cool the core).
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 103 of 110 SS4 Initiating Condition -- SITE AREA EMERGENCY Loss of all vital DC power Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss of All Vital DC power based on < 105 VDC bus voltage indications for >15 minutes.
Basis:
Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. 15 minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the event to a GENERAL EMERGENCY will occur by Abnormal Rad Levels/Radiolocial Effluent, Fission Product Barrier degradation, or Emergency Director Judgment ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 104 of 110 SS5 Initiating Condition -- SITE AREA EMERGENCY Complete loss of heat removal capability.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Heat Capacity Temperature Limit Curve exceeded.
Basis:
This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other EALs. The loss of heat removal function is indicated by the Heat Capability Temperature Limit Curve being exceeded.
Under these conditions, there is an actual major failure of a system intended for protection of the public.
Thus, declaration of a Site Area Emergency is warranted.
Escalation to General Emergency is via Abnormal Rad Levels / Radiological Effluent, Emergency Director, or Fission Product Barrier Degradation ICs.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 105 of 110 SS6 Initiating Condition -- SITE AREA EMERGENCY Inability to monitor a SIGNIFICANT TRANSIENT in progress Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. a. Loss of most or all Control Room annunciators associated with safety systems.
AND
- b. Compensatory non-alarming indications are unavailable.
AND
- c. Indications needed to monitor safety functions are unavailable.
Reactivity control Core cooling RCS status CONTAINMENT status AND
- d. SIGNIFICANT TRANSIENT is in progress.
Basis:
This IC and its associated EAL are intended to recognize the inability of the control room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.
Annunciators or indicators for this EAL include those identified the ONEPs (Off-Normal Event Procedures), and the Emergency Operating Procedures (EPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.
"Compensatory non-alarming indications" in this context includes computer based information such as SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 106 of 110 SS6(cont)
Indications needed to monitor safety functions necessary for protection of the public include control room indications, computer generated indications and dedicated annunciation capability. The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, and to maintain containment intact.
Planned and UNPLANNED actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during the transient that the cause of the loss is not an ameliorating factor.
Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an elevated risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide elevated monitoring of system operation.
There is no escalation above the Site Area Emergency for this event.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 107 of 110 SG1 Initiating Condition -- GENERAL EMERGENCY Prolonged Loss of all offsite and prolonged loss of all onsite AC power to Div I, II & III ESF busses.
Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Mode 3..............Hot Shutdown Emergency Action Level(s):
- 1. Loss of power to all of the following transformers:
ESF-11 ESF-21 ESF-12 AND Failure of Div. I, II and III Diesel Generators to supply power to emergency busses.
AND Either of the following: (a or b)
- a. Restoration of at least one emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely OR
- b. RPV level can not be maintained > - 192 in.
Basis:
Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. The four (4) hours to restore AC power is based on a site blackout coping analysis performed in conformance with 10CFR50.63 and Regulatory Guide 1.155, "Station Blackout". Appropriate allowance for offsite emergency response including evacuation of surrounding areas should be considered. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 108 of 110 SG1(cont)
This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.
In addition, under these conditions, fission product barrier monitoring capability may be degraded.
Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:
- 1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of fission product barriers is imminent?
- 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?
Thus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 109 of 110 SG3 Initiating Condition -- GENERAL EMERGENCY Failure of the reactor protection system to complete an automatic scram and manual scram was NOT successful and there is indication of an extreme challenge to the ability to cool the core Operating Mode Applicability: Mode 1..............Power Operation Mode 2..............Startup Emergency Action Level(s):
- 1. Indication(s) exist that indicate that Reactor Protection System (RPS) setpoint was exceeded and RPS automatic scram and a manual scram or ARI fails to reduce reactor power to < 4%.
AND Either of the following: (a or b)
- a. Indication(s) exists that the core cooling is extremely challenged.
- b. Indication(s) exists that heat removal is extremely challenged.
RPV pressure and suppression pool temperature cannot be maintained in the Heat Capacity Temperature Limit (HCTL) Safe Zone.
Basis:
Automatic and manual scram are not considered successful if action away from the reactor control console is required to scram the reactor.
Under the conditions of this IC and its associated EAL, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as standby liquid control and actions to insert control rods per EP-2A, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence.
The extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level as described in the EOP bases.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment II Page 110 of 110 SG3(cont)
Another consideration is the inability to initially remove heat during the stages of this sequence.
Considerations include inability to remove heat via the main condenser, or via the suppression pool.
In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (APRM downscale) a core melt sequence could exist. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment III Page 1 of 2 GUIDELINES TO TERMINATE EMERGENCY PURPOSE: To establish general guidelines to be followed should changing plant conditions warrant termination of an emergency classification.
NOTE The Emergency Director/Offsite Emergency Coordinator must discuss existing offsite conditions with appropriated State officials prior to terminating an emergency.
I. Termination Guidelines A. General
- 1. Conditions which caused the event have been terminated.
- 2. Circumstances which have arisen from the event are under control and the results of any and all pertinent data are evaluated.
- 3. All probability of recurrence of an event is removed, isolated or under control.
B. Specific Examples CATEGORY TERMINATION GUIDELINES Fires Removal/separation of any element of fire triangle. Fire under control/not spreading Spill Tanks, pipes, valves, any other problem sources are empty, isolated, and out of service.
Airborne Source identified and isolated and/or contained. Area controlled.
Explosion Existing and potential hazards removed, destroyed and/or isolated.
Abnormal Effluent Liquid discharge is terminated, sampling is completed, and statistics verified. Public exposure to Offsite radioactive material is reduced or eliminated.
Airborne - Source identified and analysis complete.
Release is terminated and its cause is under control.
All Onsite and Offsite monitoring data is evaluated.
Public exposure to Offsite radioactive material is reduced or eliminated Control Room Plant in normal emergency shutdown from remote stations. Cause Evacuation of evacuation identified and under control. No radiological conditions exist which cause the Control Room to become uninhabitable Plant Shutdown Unit is shut down by normal or emergency means. Unit is in cold Functions (not shutdown and there is no potential for uncontrolled criticality.
available or failed)
GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision xx Attachment III Page 2 of 2 CATEGORY TERMINATION GUIDELINES Fuel Handling Accident Fuel elements, segments, pellets not in a critical configuration.
- New or Spent Fuel Airborne activity has been evaluated and accountability of Damage, Channeled or components complete.
Unchanneled Water Loss - LOCA Source of water loss is defined. Ability to restore or maintain water Abnormal Primary level adequate for proper cooling.
Coolant Leak Earthquake or Other The plant has been returned to a safe condition. Threat of Natural Disaster aftershock has passed and any damage has been evaluated as to risk, if any.
Security Threat Threat to site is terminated. Probability of recurrence has been removed, with the concurrence of Security Supervisor and State, Local and Federal Officials.
GUIDELINES TO TERMINATE EMERGENCY B. Specific Examples (Cont.)
Attachment V To GNRO-2004/00057 NEI 99-01, Rev. 4 To Plant Specific Correlations, Differences, Deviations, and Justifications
GRAND GULF NUCLEAR STATION DEVIATIONS AND DIFFERENCES FROM NEI 99-01, REV 4 EMERGENCY ACTION LEVELS
GGNS NEI EAL Deviations and Differences Page 1 of 87 Introduction This document presents the GGNS site-specific deviations and differences from the NEI 99-01, Revision 4 Emergency Action Levels (EALs)
GGNS used the following definitions when determining the categorization of differences between the NEI 99-01, Revision 4 ICs abd Example EALs and the proposed GGNS ICs and EALs.
DIFFERENCE - A Difference is an EAL change where the basis scheme guidance (NUREG, NUMARC, NEI-99-01) differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the site-specific proposed EAL. Examples of differences include the use of site-specific terminology or administrative reformatting of site-specific EALs.
DEVIATION - A Deviation is an EAL change where the basis scheme guidance (NEI-99-01) differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the site-specific proposed EAL. Examples of deviations include the use of altered mode applicability, altering key words or time limits, or changing words of physical reference (protected area, safety-related equipment, etc.)
The following differences are generic in nature and apply throughout the proposed GGNS EALs:
- 1. GGNS uses formatting such as ALL CAPS, bold and underline to aid the user in applying these EALs; particularly to set apart units, time frames or quality of a value or data (such as the term valid). Formatting choices may also involve minor grammatical differences between the GGNS EALs and NEI 99-01 such as that exceeds vice exceeding, use of If, then statements for conditional statements, or the use of symbols (>, <). Such formatting differences between the GGNS EALs and NEI 99-01 will not be noted in this document as differences or deviations when they represent format choices alone and do not change the intent or materially change the content of NEI 99-01 Initiating Conditions or EALs.
- 2. At GGNS, the terms Notification of Unusual Event, NOUE, NUE, Unusual Event and UE are used interchangeably.
GGNS NEI EAL Deviations and Differences Page 2 of 87
- 3. At GGNS, all Radiological Effluent Technical Specifications or Technical Requirements Manual are included in the ODCM, thus ODCM is used in place of Radiological Effluent Technical Specifications references.
- 4. Safeguards Contingency Plan is the term used to encompass all security plans/documents.
- 5. Reference to emergency generator(s) in EALs was changed to emergency diesel generator(s) for consistency with site terminology.
- 6. RPV inventory and RCS inventory is used inconsistently in the NEI document. GGNS used RCS when applicable or referring to inventory and RPV when applicable or referring to a level. The term RPV level is consistent with the GGNS EPs and SAPs. The intent and meaning of the EALs are maintained.
- 7. Technical Specification cold shutdown temperature limit is equal to 200 °F.
- 9. The terms increase and decrease have been replaced with equivalent terms in accordance with GGNS standard practices.
- 10. Differences in Operating modes:
NEI 99-01 Operating Modes: Power Operation Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled Grand Gulf Operating Modes: Power Operation (1)
Startup (2)
Hot Shutdown (3)
Cold Shutdown (4)
Refueling (5)
Defueled (D)
GGNS NEI EAL Deviations and Differences Page 3 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or Longer.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
- 1. VALID reading on any effluent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
- 2. VALID reading on one or more of the following radiation monitors that exceeds the reading shown for 60 minutes or longer:
(site-specific list)
- 3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 60 minutes or longer, in excess of two times (site-specific technical specifications).
- 4. VALID reading on perimeter radiation monitoring system greater than 0.10 mR/hr above normal background sustained for 60 minutes or longer [for sites having telemetered perimeter monitors].
- 5. VALID indication on automatic real-time dose assessment capability greater than (site-specific value) for 60 minutes or longer [for sites having such capability].
Differences:
- 1. EAL #1, the setpoint is established by the ODCM to alarm at 50% of the TRM limit. Thus four times the setpoint corresponds to two times the TRM limit.
Deviations:
- 1. GGNS has no perimeter radiation monitoring system, thus EAL #4 of NEI 99-01 Rev. 4 is not applicable.
GGNS NEI EAL Deviations and Differences Page 4 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Unexpected Increase in Plant Radiation.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. a. VALID (site-specific) indication of uncontrolled water level decrease in the reactor refueling cavity, spent fuel pool, or fuel transfer canal with all irradiated fuel assemblies remaining covered by water.
AND
- b. Unplanned VALID (site-specific) Direct Area Radiation Monitor reading increases
- 2. Unplanned VALID Direct Area Radiation Monitor readings increases by a factor of 1000 over normal* levels.
- Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
Differences:
GGNS terminology uses area radiation monitors instead of direct area radiation monitors.
- 2. Reworded EAL #1 for GGNS terminology (e.g., Upper Ctmt. Pools instead of reactor refueling cavity, and Aux Bldg Fuel Pools instead of spent fuel pool).
- 3. Some of the area Radiation Monitors installed at Grand Gulf do not have sufficient range to allow the determination of an increase by a factor of 1000. Therefore, for these ARMs full scale is conservatively used in lieu of the factor of 1000.
- 4. The note for normal levels is contained in the EAL and not as a separate note. This is considered a difference since the wording agrees with the NEI meaning and intent.
Deviations:
- 1. In GGNS EAL #1 site specific level indicators are not included for pool levels. Guidance for methods of determining lowering level is included in
GGNS NEI EAL Deviations and Differences Page 5 of 87 the bases and not the EAL body, since site specific level indicators are not available. This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 6 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA1 Initiating Condition -- ALERT Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or Longer.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
- 1. VALID reading on any effluent monitor that exceeds 200 times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer.
- 2. VALID reading on one or more of the following radiation monitors that exceeds the reading shown for 15 minutes or longer:
(site-specific list)
- 3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 15 minutes or longer, in excess of 200 times (site-specific technical specifications).
- 4. VALID reading on perimeter radiation monitoring system greater than 10.0 mR/hr above normal background sustained for 15 minutes or longer [for sites having telemetered perimeter monitors].
- 5. VALID indication on automatic real-time dose assessment capability greater than (site-specific value) for 15 minutes or longer [for sites having such capability].
Differences:
- 1. EAL #1, the setpoint is established by the ODCM to alarm at 50% of the TRM limit. Thus four hundred times the setpoint corresponds to two hundred times the TRM limit.
Deviations:
- 1. GGNS has no perimeter radiation monitoring system, thus EAL #4 of NEI 99-01 Rev. 4 is not applicable.
4 is not applicable.
GGNS NEI EAL Deviations and Differences Page 7 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA2 Initiating Condition -- ALERT Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. A VALID (site-specific) alarm or reading on one or more of the following radiation monitors: (site-specific monitors)
Refuel Floor Area Radiation Monitor Fuel Handling Building Ventilation Monitor Refueling Bridge Area Radiation Monitor
- 2. Water level less than (site-specific) feet for the reactor refueling cavity, spent fuel pool and fuel transfer canal that will result in irradiated fuel uncovering.
Differences:
- 1. Reworded EAL #2 for GGNS terminology (e.g., Upper Ctmt. Pools instead of reactor refueling cavity, and Aux Bldg Fuel Pools instead of spent fuel pool).
Deviations:
- 1. Grand Gulf has no direct water level instrumentation in the affected areas.
Thus a confirmed observed decrease in water level is the most valid method to determine irradiated fuel uncovery.
GGNS NEI EAL Deviations and Differences Page 8 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA3 Initiating Condition -- ALERT Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. VALID (site-specific) radiation monitor readings GREATER THAN 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions:
(Site-specific) list
- 2. VALID (site-specific) radiation monitor readings GREATER THAN <site specific> values in areas requiring infrequent access to maintain plant safety functions.
(Site-specific) list Differences:
- 1. EAL #1 does not list the Rad Waste Control Room or the Central Alarm Station for the following reasons. The Rad Waste Control Room is not required to be continuously manned and the Central Alarm Station is located within the Main Control Room envelope.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 9 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AS1 Initiating Condition -- SITE AREA EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the Actual or Projected Duration of the Release.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4)
Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently escalated.
- 1. VALID reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:
(site-specific list)
- 2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE or 500 mR thyroid CDE at or beyond the site boundary.
- 3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 100 mR/hr. [for sites having telemetered perimeter monitors]
- 4. Field survey results indicate closed window dose rates exceeding 100 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 500 mR for one hour of inhalation, at or beyond the site boundary.
Differences:
Deviations:
- 1. EAL #3 of NEI 99-01 Rev. 4 was not used at GGNS. GGNS has no perimeter radiation monitoring system.
GGNS NEI EAL Deviations and Differences Page 10 of 87 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AG1 Initiating Condition -- GENERAL EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4)
Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently escalated.
- 1. VALID reading on one or more of the following radiation monitors that exceeds or expected to exceed the reading shown for 15 minutes or longer:
(site-specific list)
- 2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE or 5000 mR thyroid CDE at or beyond the site boundary.
- 3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 1000 mR/hr. [for sites having telemetered perimeter monitors]
- 4. Field survey results indicate closed window dose rates exceeding 1000 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 5000 mR for one hour of inhalation, at or beyond site boundary.
Differences:
- 1. The wording associated with the NOTE was reworded to address the fact that a General Emergency can not be escalated.
GGNS NEI EAL Deviations and Differences Page 11 of 87 Deviations:
- 1. EAL #3 of NEI 99-01 Rev. 4 was not used at GGNS. GGNS has no perimeter radiation monitoring system.
GGNS NEI EAL Deviations and Differences Page 12 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Reactor Coolant System Leakage Operating Mode Applicability: Cold Shutdown Emergency Action Levels: (1 or 2)
- 1. Unidentified or pressure boundary leakage greater than 10 gpm.
- 2. Identified leakage greater than 25 gpm.
Differences:
None Deviations:
- 1. NEI 99-01 was not used as written in the GGNS EALs. Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit. Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the T.S. 18 month surveillance tests since this is the only time the drywell is accessible.
The sump tanks, pumps, check valves and level switches are in the drywell.
An EAL threshold value was established using RPV water level (-41.6 in.)
instead of a Leak Detection System leakage rate to provide the operator with a valid and available indication of the status of RCS inventory. The RPV level of -41.6 in. is ~ 115 inches above the top of active fuel and is ~ 108 inches above the RCS leakage ALERT threshold. In Mode 4, the level indication in the control room is on the control panel (P601) that has the Tech Spec required ECCS controls and indications. The reduction of RPV level would alert the operator to evaluate and correct the situation.
RHR shutdown cooling, ADHR (Alternate Decay Heat Removal), and RWCU are the systems normally operated in Mode 4 that have a potential for an RCS leak. The RHR shutdown cooling mode and ADHR are designed to isolate the suction header MOVs if leakage is indicated by lowering RPV level. The level instrumentation for isolation is required in Mode 4 per T.S. 3.3.6.1. If an uncontrolled water level drop is occurring, the potential RCS leakage paths will isolate the system when RPV level drops to 11.4 in. If level continues to lower, the T.S. required ECCS system or
GGNS NEI EAL Deviations and Differences Page 13 of 87 other available systems would be used for injection. The Primary Containment Isolation Valve isolation will occur if the leakage is from RWCU outside of containment. The use of RPV level is consistent with the ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY classifications for RCS leakage. A lowering of the RPV level to -41.6 in. will be considered a potential degradation of the level of safety of the plant and an UNUSUAL EVENT will be declared. This meets the intent of an UNUSUAL EVENT as defined in NEI 99-01.
The use of RPV level indication in the EAL instead of a leakage rate meets the intent of the NEI EAL as described in Appendix C. Appendix C states that shutdown events are based on the potential loss or loss of one or more of the cold shutdown barrier functions. The concern for the shutdown EAL is the inadvertent draining of the reactor vessel and loss of residual heat removal capability as based on industry operating experience. The NEI EAL qualifiers of 10 gpm unidentified leakage or 25 gpm total leakage do not represent a potential to drain the reactor vessel or cause a loss of heat removal with the reactor in cold shutdown. Since the RPV is not at pressure, propagation is not expected to reach the critical crack size to drain the vessel. Additionally, any leakage detected in Mode 4, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must be analyzed to make the determination that the leak is from the RCS inventory.
Based on the above, unidentified leakage and total leakage are not valid indications of RCS leakage and would not be used to declare an emergency.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 14 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of RCS inventory with irradiated fuel in the reactor vessel Operating Mode Applicability: Refueling Emergency Action Levels: (1 or 2)
- 2. a. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level increase AND
- b. RPV level cannot be monitored Differences:
None Deviations:
- 1. NEI 99-01 EAL #2.a was not used as written in the GGNS EALs. Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit.
Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the T.S. 18 month surveillance tests since this is the only time the drywell is accessible. The sump tanks, pumps, check valves and level switches are in the drywell.
The use of words unexpected loss of RCS inventory instead of the use of the qualifier as indicated by unexplained {site specific} sump and tank level increase meets the intent of the NEI EAL. If RPV level indication is lost, any means of detecting a loss of RCS inventory should be utilized.
Additionally, any leakage detected in Mode 4, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must be analyzed to make the determination that the leak is from the RCS inventory.
GGNS NEI EAL Deviations and Differences Page 15 of 87 Based on the above, allowing for a more diverse means of detecting unexpected RCS inventory loss is justified.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 16 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU3 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of all offsite power to Essential busses for greater than 15 minutes.
Operating Mode Applicability: Cold Shutdown Refueling Emergency Action Levels:
- 1. a. Loss of power to (site-specific) transformers for greater than 15 minutes.
AND
- b. At least (site-specific) emergency generators are supplying power to emergency busses.
Differences:
- 1. Initiating Condition of CU3 in GGNSs EALs was reworded to use Div. I & II ESF instead of essential.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 17 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU4 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel.
Operating Mode Applicability: Cold Shutdown Refueling Emergency Action Levels: (1 or 2)
- 1. An UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 18 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU5 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Fuel clad degradation Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)
- 1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.
- 2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.
Differences:
None Deviations:
The Main Steam Line radiation monitors and Offgas pre-treatment radiation monitors are the available instruments to monitor for fuel clad degradation in Modes 1, 2, and 3. The instruments are not available in Modes 4 or 5.
These instruments are not reliable indications of clad damage in Modes 4 and 5 since steam or coolant would normally not be present in the Main Steam Lines outside the Containment, nor in Offgas. Therefore, this EAL is not applicable.
The Abnormal Radiation / Effluent EALs are applicable in all modes and meet the intent of detection of fuel clad degradation and potential degradation in the level of safety of the plant for this EAL. Abnormal Radiation level EAL AU1 (Any unplanned release of gaseous or liquid radioactivity to the environment exceeds 2 X ODCM limit for 60 minutes),
AU2 (Unplanned release in plant area radiation levels by a factor of 1000) or AA3 (Damage to irradiated fuel or unplanned loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel) are precursors to alert personnel of radiological levels of a magnitude to indicate fuel clad damage.
GGNS NEI EAL Deviations and Differences Page 19 of 87 In Mode 4 and 5, the loss of RCS inventory below the top of active fuel would be the mechanism for potential fuel clad damage. If this event occurred, a site area emergency CS1 would be applicable.
In mode 5, a fuel handling event is the most probable mechanism for potential fuel clad damage. If this event occurred, an alert emergency AA3 would be applicable.
This deviation does not reduce the effectiveness and does not impact the health and safety of the public since other indicators are available to determine if an EAL is necessary.
This EAL addresses coolant samples exceeding coolant technical specifications for iodine spike. The GGNS Technical Specification (3.4.8) for RCS activity is not applicable in Modes 4 and 5 or in Modes 2 and 3 if the main steam lines isolated. This license amendment deleted the requirement for coolant samples for specific activity in all operation modes except Mode 1 and in Modes 2 and 3 when the main steam lines are not isolated.
The TS bases for the RCS specific activity limit is to ensure, in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10CFR50.67 limit. The associated action statement for exceeding the TS limit in Modes 1, 2, or 3 is to restore limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or shutdown to Mode 4 if time limit is exceeded or specific activity is >4 µCi/gm dose equivalent I-131. There is no required actions after Mode 4 or 5 is obtained and coolant activity is >4
µCi/gm dose equivalent I-131. If an EAL were applicable in this instance, the declaration of a NOUE would be required even though the site is in compliance with TS.
The T.S. basis for the limit on the level of radioactivity in the reactor coolant is to ensure radiation doses are maintained within the limits of 10CFR50.67 in the event of a release of any radioactive material to the environment during a Design Basis Accident (DBA). The release of coolant during a DBA in modes 1, 2, and 3 could send radioactive materials into the environment.
The limits on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses at the site boundary, resulting from a Main Steam Line Break outside containment during steady state operation, will not exceed 10% of the dose guidelines of 10CFR50.67. These accidents are not analyzed for Modes 4 and 5.
In modes 4 and 5, fuel clad degradation will not occur without some initiating event to cause clad damage. Other EALS are in place for such events or indications of fuel clad degradation. The Abnormal Radiation / Effluent EALs
GGNS NEI EAL Deviations and Differences Page 20 of 87 are applicable in all modes and meet the intent of detection of fuel clad degradation and potential degradation in the level of safety of the plant for this EAL. Abnormal Radiation level EAL AU1 (Any unplanned release of gaseous or liquid radioactivity to the environment exceeds 2 X ODCM limit for 60 minutes), AU2 (Unplanned release in plant area radiation levels by a factor of 1000) or AA3 (Damage to irradiated fuel or unplanned loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel) are precursors to alert personnel of radiological levels of a magnitude to indicate fuel clad damage.
In Mode 4 and 5, the loss of RCS inventory below the top of active fuel would be the mechanism for potential fuel clad damage. If this event occurred, a site area emergency CS1 would be applicable.
In mode 5, a fuel handling event is the most probable mechanism for potential fuel clad damage. If this event occurred, an alert emergency AA3 would be applicable and a coolant sample for specific activity to determine clad damage for a NOUE would not be required.
Section 3.7 of NEI 99-01 states that an Unusual Event indicate a potential degradation of the level of safety of the plant and that potential degradation of the level of safety is indicated primarily by exceeding plant technical specification Limiting Condition of Operation (LCO) allowable action time for achieving required mode change.
Section 5.2 of NEI 99-01 states that the primary threshold for NOUEs as operation outside the safety envelope for the plant as defined by plant technical specifications, including LCOs and Action Statement times.
Based on the above factors, other EALs are in place for emergency declaration in events causing fuel clad degradation. Technical Specifications for coolant activity are not applicable in Modes 4 and 5. Therefore the EAL for reactor coolant sample activity in Modes 4 and 5 do not apply. This deviation does not reduce the effectiveness and does not impact the health and safety of the public since other EALs would be considered for this condition.
GGNS NEI EAL Deviations and Differences Page 21 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU6 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.
Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)
- 1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.
- 2. Loss of all (site-specific list) offsite communications capability.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 22 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU7 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Required DC Power for Greater than 15 Minutes.
Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Level:
- 1. a. UNPLANNED Loss of Vital DC power to required DC busses based on (site-specific) bus voltage indications.
AND
- b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 23 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.
Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)
- 1. An UNPLANNED extended positive period observed on nuclear instrumentation.
- 2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
Differences:
Deviations:
GGNS NEI EAL Deviations and Differences Page 24 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA1 Initiating Condition -- ALERT Loss of RCS Inventory.
Operating Mode Applicability: Cold Shutdown Example Emergency Action Levels: (1 or 2)
(low-low ECCS actuation setpoint) (BWR)
(bottom ID of the RCS loop) (PWR)
- 2. a. Loss of RCS inventory as indicated by unexplained {site-specific} sump and tank level increase AND
- b. RCS level cannot be monitored for > 15 minutes Differences:
None Deviations:
- 1. NEI 99-01 EAL #2.a was not used as written in the GGNS EALs. Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit.
Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the T.S. 18 month surveillance tests since this is the only time the drywell is accessible. The sump tanks, pumps, check valves and level switches are in the drywell.
The use of words unexpected loss of RCS inventory instead of the use of the qualifier as indicated by unexplained {site specific} sump and tank level increase meets the intent of the NEI EAL. If RPV level indication is lost, any means of detecting a loss of RCS inventory should be utilized.
Additionally, any leakage detected in Mode 4, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must be analyzed to make the determination that the leak is from the RCS inventory.
GGNS NEI EAL Deviations and Differences Page 25 of 87 Based on the above, allowing for a more diverse means of detecting unexpected RCS inventory loss is justified.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 26 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA2 Initiating Condition -- ALERT Loss of RPV Inventory with Irradiated Fuel in the RPV.
Operating Mode Applicability: Refueling Example Emergency Action Levels: (1 or 2)
(low-low ECCS actuation setpoint) (BWR)
(bottom ID of the RCS loop) (PWR)
- 2. a. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level increase AND
- b. RPV level cannot be monitored for > 15 minutes Differences:
None Deviations:
- 1. NEI 99-01 EAL #2.a was not used as written in the GGNS EALs. Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit.
Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the T.S. 18 month surveillance tests since this is the only time the drywell is accessible. The sump tanks, pumps, check valves and level switches are in the drywell.
The use of words unexpected loss of RCS inventory instead of the use of the qualifier as indicated by unexplained {site specific} sump and tank level increase meets the intent of the NEI EAL. If RPV level indication is lost, any means of detecting a loss of RCS inventory should be utilized.
Additionally, any leakage detected in Mode 5, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must
GGNS NEI EAL Deviations and Differences Page 27 of 87 be analyzed to make the determination that the leak is from the RCS inventory.
Based on the above, allowing for a more diverse means of detecting unexpected RCS inventory loss is justified.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 28 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA3 Initiating Condition -- ALERT Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.
Operating Mode Applicability: Cold Shutdown Refueling Defueled Example Emergency Action Level:
- 1. a. Loss of power to (site-specific) transformers.
AND
- b. Failure of (site-specific) emergency generators to supply power to emergency busses.
AND
- c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.
Differences:
- 2. Initiating Condition of CA3 in GGNSs EALs was reworded to use Div. I & II ESF instead of essential.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 29 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA4 Initiating Condition -- ALERT Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV.
Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (EAL 1 or 2 or 3)
- 1. With CONTAINMENT CLOSURE and RCS integrity not established an UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit.
- 2. With CONTAINMENT CLOSURE established and RCS integrity not established or RCS inventory reduced an UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 20 minutes.
- 3. An UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 60 minutes or results in an RCS pressure increase of greater than {site specific} psig.
Differences:
Deviations:
- 1. Reduced Inventory is not applicable to GGNS.
GGNS NEI EAL Deviations and Differences Page 30 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1 Initiating Condition -- SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability.
Operating Mode Applicability: Cold Shutdown Example Emergency Action Levels: (1 or 2)
- 1. With CONTAINMENT CLOSURE not established:
(6" below the low-low ECCS actuation setpoint)
(BWR)
(6" below the bottom ID of the RCS loop)
(PWR)
- b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level increase
- 2. With CONTAINMENT CLOSURE established
- b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as indicated by either:
- Unexplained {site-specific} sump and tank level increase
- Erratic Source Range Monitor Indication Differences:
None Deviations:
Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit. Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the T.S. 18 month surveillance tests since this is the only time the drywell is accessible. The sump tanks, pumps, check valves and level switches are in the drywell.
GGNS NEI EAL Deviations and Differences Page 31 of 87 The use of words unexpected loss of RCS inventory instead of the use of the qualifier as indicated by unexplained {site specific} sump and tank level increase meets the intent of the NEI EAL. If RPV level indication is lost, any means of detecting a loss of RCS inventory should be utilized.
Additionally, any leakage detected in Mode 4, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must be analyzed to make the determination that the leak is from the RCS inventory.
Based on the above, allowing for a more diverse means of detecting unexpected RCS inventory loss is justified.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 32 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS2 Initiating Condition -- SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV.
Operating Mode Applicability: Refueling Example Emergency Action Levels: (1 or 2)
- 1. With CONTAINMENT CLOSURE not established:
(6" below the low-low ECCS actuation setpoint) (BWR)
(6" below the bottom ID of the RCS loop) (PWR)
- b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:
- Containment High Range Radiation Monitor reading > {site-specific}
setpoint
- Erratic Source Range Monitor Indication
- Other {site-specific} indications
- 2. With CONTAINMENT CLOSURE established
- b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:
- Containment High Range Radiation Monitor reading > {site-specific}
setpoint
- Erratic Source Range Monitor Indication
- Other {site-specific} indications Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 33 of 87 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CG1 Initiating Condition -- GENERAL EMERGENCY Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV.
Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Level: (1 and 2 and 3)
- 2. RPV Level:
- b. cannot be monitored with Indication of core uncovery for > 30 minutes as evidenced by one or more of the following:
- Containment High Range Radiation Monitor reading > {site-specific}
setpoint
- Erratic Source Range Monitor Indication
- Other {site-specific} indications
- 3. {Site specific} indication of CONTAINMENT challenged as indicated by one or more of the following:
- Explosive mixture inside containment
- Pressure above {site specific} value
- CONTAINMENT CLOSURE not established
- Secondary Containment radiation monitors above {site specific}
value (BWR only)
Differences:
None Deviations:
- 1. NEI 99-01 EAL #1 was not used as written in the GGNS EALs. Reactor coolant leakage monitoring is not required in operating modes 4 and 5 per GGNS Technical Specifications section 3.4.5 and there is no LCO limit.
Leakage propagation is not a concern, as at operating pressure, since the reactor is shut down and depressurized. The Leak Detection System may not be available in Modes 4 or 5 due to maintenance activities such as the
GGNS NEI EAL Deviations and Differences Page 34 of 87 T.S. 18 month surveillance tests since this is the only time the drywell is accessible. The sump tanks, pumps, check valves and level switches are in the drywell.
The use of words unexpected loss of RCS inventory instead of the use of the qualifier as indicated by unexplained {site specific} sump and tank level increase meets the intent of the NEI EAL. If RPV level indication is lost, any means of detecting a loss of RCS inventory should be utilized.
Additionally, any leakage detected in Mode 4 or 5, if the Leak Detection System instrumentation is in service, a chemistry sample of the sump contents must be analyzed to make the determination that the leak is from the RCS inventory.
Based on the above, allowing for a more diverse means of detecting unexpected RCS inventory loss is justified.
This deviation agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 35 of 87 EVENTS RELATED TO ISFSI E-HU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: Not applicable Example Emergency Action Level: (1 or 2 or 3)
- 1. Natural phenomena events affecting a loaded cask CONFINEMENT BOUNDARY.
(site-specific list)
- 2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.
(site-specific list)
- 3. Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY.
Differences:
None Deviations:
Mode applicability was changed to All for human factoring concerns during the review of the EALs. Some reviewers believed that it was more appropriate to list all modes to preclude an operator from inferring that, since mode applicability was not applicable, he would not have to declare an emergency class.
GGNS NEI EAL Deviations and Differences Page 36 of 87 EVENTS RELATED TO ISFSI E-HU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Confirmed Security Event with potential loss of level of safety of the ISFSI.
Operating Mode Applicability: Not applicable Example Emergency Action Levels:
- 1. Security Event as determined from (site-specific) Security Plan and reported by the (site-specific) security shift supervision.
Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 37 of 87 FUEL CLAD BARRIER EXAMPLE EALS
- 1. Primary Coolant Activity Level Loss: Coolant Activity greater than (site-specific) value Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 38 of 87 FUEL CLAD BARRIER EXAMPLE EALS
- 2. Reactor Vessel Water Level Loss: Level less than (site-specific) value Potential Loss Level less than (site specific) value Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 39 of 87 FUEL CLAD BARRIER EXAMPLE EALS
- 3. Drywell Radiation Monitoring Loss: Drywell radiation monitor reading > (site-specific) R/hr Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 40 of 87 FUEL CLAD BARRIER EXAMPLE EALS
- 4. Other (Site-Specific) Indications Loss: (site-specific) as applicable Potential Loss (site-specific) as applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
The procedure used in the TSC and EOF to estimate core damage provides a decision making process based on 3 parameters:
Drywell radiation levels, Containment radiation levels, and zirconium oxidation. RPV water level, time of shutdown and hydrogen concentration are considered in the decision making.
The fuel clad barrier EALs use coolant activity, RPV level and radiation level for indication of fuel failure. Use of a hydrogen concentration EAL is not needed since the initiating condition, loss of core cooling, would have already met the fuel clad failure condition before hydrogen is generated. Hydrogen concentration is not considered for an EAL for fuel clad barrier because of these factors.
After evaluating the SAPs, GGNS determined a site specific EAL was not applicable for NEI Fuel Clad Barrier EAL 4. This is not considered a deviation since additional conditions are not applicable for the EAL. The decision to exclude the EAL does not reduce the effectiveness and does not impact the health and safety of the public.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 41 of 87 FUEL CLAD BARRIER EXAMPLE EALS
- 5. Emergency Director Judgment Loss or Potential Loss: Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 42 of 87 RCS BARRIER EXAMPLE EALs:
1 Drywell Pressure Loss: Drywell pressure greater than (site-specific) psig Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
Added the qualifier with indications of a reactor coolant leak in the drywell as an aid to operators in making timely decisions. The qualifier is in the NEI bases and the intent of the EAL is met since the EAL is RCS barrier. The addition of this qualifier agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 43 of 87 RCS BARRIER EXAMPLE EALs
- 2. Reactor Vessel Water Level Loss: Level less than (site-specific value)
Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
Added the qualifier with indications of a reactor coolant leak in the drywell as an aid to operators in making timely decisions. The qualifier is in the NEI bases and the intent of the EAL is met since the EAL is RCS barrier. The addition of this qualifier agrees with the NEI meaning and intent, does not reduce the effectiveness and does not impact the health and safety of the public.
GGNS NEI EAL Deviations and Differences Page 44 of 87 RCS BARRIER EXAMPLE EALs
- 3. RCS Leak Rate Loss: (site-specific) indication of an unisolable steam line break Potential Loss: RCS leakage greater than 50 gpm inside the drywell Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 45 of 87 RCS BARRIER EXAMPLE EALs
- 4. Drywell Radiation Monitoring Loss: Drywell radiation monitor reading greater than (site-specific) R/hr Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 46 of 87 RCS BARRIER EXAMPLE EALs
- 5. Other (Site-Specific) Indications Loss: (site-specific) as applicable Potential loss: (site-specific) as applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. This EAL was not implemented at GGNS because there are no other site-specific indicators available for this EAL. The EOP and SAP procedures were reviewed for additional indications of RCS leakage. No additional methods were identified.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 47 of 87 RCS BARRIER EXAMPLE EALs
- 6. Emergency Director Judgment Loss: Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Potential loss: Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Applicability: Power Operation, Startup, Hot Shutdown Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 48 of 87 CONTAINMENT BARRIER EXAMPLE EALs:
- 1. Drywell Pressure Loss: Rapid unexplained decrease following initial increases OR Drywell pressure response not consistent with LOCA conditions Potential Loss: (site-specific) psig and increasing OR Explosive mixture exists Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. Containment pressure is used instead of Drywell pressure. Grand Gulf has a Mark III containment design. In the Mark III containment design, the Containment is the fission product barrier. The use of the term Primary Containment is consistent with T.S and EOPs. The intent of the EAL is still met.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 49 of 87 CONTAINMENT BARRIER EXAMPLE EALs
- 2. Reactor Vessel Water Level Loss: Not applicable Potential Loss: Primary Containment flooding required Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. GGNS uses the term Primary Containment as used in Tech Specs - intent of EAL is met Deviations:
None
GGNS NEI EAL Deviations and Differences Page 50 of 87 CONTAINMENT BARRIER EXAMPLE EALs
- 3. Containment Isolation Failure or Bypass Loss: Failure of both valves in any one line to close AND downstream pathway to the environment exists OR Intentional venting per EOPs OR Unisolable primary system leakage outside drywell as indicated by area temperature or area radiation alarm Potential Loss: Not applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. The Grand Gulf EAL references primary containment venting required by the Severe Accident Procedures (SAP) or the Emergency Operating Procedures (EOP). This is considered a difference instead of a deviation since the intent of the EAL is met.
- 2. GGNS uses the term Primary Containment as used in Tech Specs and EOPs - intent of EAL is met
None
GGNS NEI EAL Deviations and Differences Page 51 of 87 CONTAINMENT BARRIER EXAMPLE EALs
- 4. Significant Radioactive Inventory in Containment Loss: Not applicable Potential Loss: Drywell radiation monitor reading greater than (site-specific) R/hr Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. GGNS uses the term Primary Containment as used in Tech Specs - intent of EAL is met.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 52 of 87 CONTAINMENT BARRIER EXAMPLE EALs
- 5. Other (Site-Specific) Indications Loss: (site-specific) as applicable Potential Loss: (site-specific) as applicable Applicability: Power Operation, Startup, Hot Shutdown Differences:
- 1. This EAL is not used since additional conditions are not specified. Other conditions in the EOPs and SAPs that may represent a challenge to the containment are the unsafe zones in the Pressure Suppression curve and the Primary Containment Pressure Limit curve. Both conditions are concerns when in containment flooding per the severe accident procedure (SAP).Both curves are based on containment pressure and suppression pool/containment water level. These parameters were not included as a challenge to containment. A NOUE would have already been declared on the entry into the SAP.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 53 of 87 CONTAINMENT BARRIER EXAMPLE EALs
- 6. Emergency Director Judgment Loss: Any condition in the opinion of the Emergency Director that indicates a loss of the Primary Containment barrier Potential Loss: Any condition in the opinion of the Emergency Director that indicates a potential loss of the Primary Containment barrier Applicability: Power Operation, Startup, Hot Shutdown Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 54 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Natural and Destructive Phenomena Affecting the PROTECTED AREA.
Operating Mode Applicability: All Example Emergency Action Level: (1 or 2 or 3 or 4 or 5 or 6 or 7)
- 1. (Site-Specific) method indicates felt earthquake.
- 2. Report by plant personnel of tornado or high winds greater than (site-specific) mph striking within PROTECTED AREA boundary.
- 3. Vehicle crash into plant structures or systems within PROTECTED AREA boundary.
- 4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.
- 5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 6. Uncontrolled flooding in (site-specific) areas of the plant that has the potential to affect safety related equipment needed for the current operating mode.
- 7. (Site-Specific) occurrences affecting the PROTECTED AREA.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 55 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutes of Detection.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. FIRE in buildings or areas contiguous to any of the following (site-specific) areas not extinguished within 15 minutes of control room notification or verification of a control room alarm:
(Site-specific) list Differences:
NEI 99-01 Rev. 4 HU2 was renumbered to HU4 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 56 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect NORMAL PLANT OPERATIONS.
- 2. Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 57 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. Security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision
- 2. A credible site specific security threat notification.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 58 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU5 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Differences:
NEI 99-01 Rev. 4 HU5 was renumbered to HU2 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 59 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition -- ALERT Natural and Destructive Phenomena Affecting the Plant VITAL AREA.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5 or 6)
- 1. (Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake (OBE).
- 2. Tornado or high winds greater than (site-specific) mph within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures / equipment or Control Room indication of degraded performance of those systems.
- Reactor Building
- Intake Building
- Refueling Water Storage Tank
- Diesel Generator Building
- Turbine Building
- Condensate Storage Tank
- Control Room
- Other (Site-Specific) Structures.
- 3. Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures or equipment therein or control indication of degraded performance of those systems:
- Reactor Building
- Intake Building
- Refueling Water Storage Tank
- Diesel Generator Building
- Turbine Building
- Condensate Storage Tank
- Control Room
- Other (Site-Specific) Structures.
- 4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of the following plant areas: (site-specific) list.
GGNS NEI EAL Deviations and Differences Page 60 of 87
- 5. Uncontrolled flooding in (site-specific) areas of the plant that results in degraded safety system performance as indicated in the control room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.
- 6. (Site-Specific) occurrences within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to plant structures containing equipment necessary for safe shutdown, or has caused damage as evidenced by control room indication of degraded performance of those systems.
Differences:
- 2. GGNS Table H2 list of site specific structures that contain systems and functions for safe shutdown does not include all areas listed in NEI HA1 EALs #2 and 3 as instructed in the NEI bases document.
- 3. GGNS Table H2 list of site specific structures that contain systems and functions for safe shutdown is used for EAL #4. Even though some structures may not be at risk for penetration from a turbine generated missile, the same list was used for consistency in the EALs.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 61 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating Condition -- ALERT FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. FIRE or EXPLOSION in any of the following (site-specific) areas:
(Site-specific) list AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.
Differences:
NEI 99-01 Rev. 4 HA2 was renumbered to HA4 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 62 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3 Initiating Condition -- ALERT Release of Toxic or Flammable Gases Within or Contiguous to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. Report or detection of toxic gases within or contiguous to a VITAL AREA in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).
- 2. Report or detection of gases in concentration greater than the LOWER FLAMMABILITY LIMIT within or contiguous to a VITAL AREA.
Differences:
NEI 99-01 Rev. 4 HA3 was renumbered to HA5 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 63 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4 Initiating Condition -- ALERT Confirmed Security Event in a Plant PROTECTED AREA.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE.
- 2. Other security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 64 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 Initiating Condition -- ALERT Control Room Evacuation Has Been Initiated.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Entry into (site-specific) procedure for control room evacuation.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 65 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6 Initiating Condition -- ALERT Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Differences:
NEI 99-01 Rev. 4 HA6 was renumbered to HA2 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 66 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition - SITE AREA EMERGENCY Confirmed Security Event in a Plant VITAL AREA.
Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)
- 1. INTRUSION into the plant VITAL AREA by a HOSTILE FORCE.
- 2. Other security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 67 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition - SITE AREA EMERGENCY Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Control room evacuation has been initiated.
AND Control of the plant cannot be established per (site-specific) procedure within (site-specific) minutes.
Differences:
NEI 99-01 Rev. 4 HS2 was renumbered to HS3 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 68 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS3 Initiating Condition - SITE AREA EMERGENCY Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for 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.
Differences:
NEI 99-01 Rev. 4 HS3 was renumbered to HS2 in GGNSs EALs for formatting purposes.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 69 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 Initiating Condition - GENERAL EMERGENCY Security Event Resulting in Loss Of Physical Control of the Facility.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.
Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 70 of 87 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG2 Initiating Condition - GENERAL EMERGENCY Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency.
Operating Mode Applicability: All Example Emergency Action Level:
- 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 71 of 87 SYSTEM MALFUNCTION SU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of All Offsite Power to essential Busses for Greater Than 15 Minutes.
Operating Mode Applicability: Power Operation (1)
Startup (2)
Hot Standby (3)
Hot Shutdown (4)
Example Emergency Action Level:
- 1. Loss of power to (site-specific) transformers for greater than 15 minutes.
AND At least (site-specific) emergency generators are supplying power to emergency busses.
Difference:
- 1. Initiating Condition of SU1 in GGNSs EALs was reworded to use Div. I & II ESF instead of essential.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 72 of 87 SYSTEM MALFUNCTION SU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inability to Reach Required Shutdown Within Technical Specification Limits.
Operating Mode Applicability: Power Operation (1)
Startup (2)
Hot Standby (3)
Hot Shutdown (4)
Example Emergency Action Level:
- 1. Plant is not brought to required operating mode within (site-specific)
Technical Specifications LCO Action Statement Time.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 73 of 87 SYSTEM MALFUNCTION SU3 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with safety systems for greater than 15 minutes.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 74 of 87 SYSTEM MALFUNCTION SU4 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Fuel Clad Degradation.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)
- 1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.
- 2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 75 of 87 SYSTEM MALFUNCTION SU5 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT RCS Leakage.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)
- 1. Unidentified or pressure boundary leakage greater than 10 gpm.
- 2. Identified leakage greater than 25 gpm.
Differences:
Deviations:
- 1. The Grand Gulf Technical Specification limit is 5 gpm for unidentified leakage and 30 gpm total leakage averaged over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NEI EAL for identified leakage is less than the allowable TS, therefore the GGNS EAL value for EAL #2 is 35 gpm to be above allowable limits The required TS LCO for 30 gpm identified leakage is to reduce leakage to within the TS limit of 30 gpm within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If a leakage value were between the TS 30 gpm and the EAL 35 gpm and the LCO condition could not be met, the event would be declared on SU11. If leakage cannot be reduced to within the TS allotted time and reaches 35 gpm, SU 7 EAL will be applicable.
GGNS NEI EAL Deviations and Differences Page 76 of 87 SYSTEM MALFUNCTION SU6 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)
- 1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.
- 2. Loss of all (site-specific list) offsite communications capability.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 77 of 87 SYSTEM MALFUNCTION SU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.
OPERATING MODE APPLICABILITY Hot Standby Hot Shutdown Example Emergency Action Level: (1 or 2)
- 1. An UNPLANNED extended positive period observed on nuclear instrumentation.
- 2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.
Differences:
Deviations:
GGNS NEI EAL Deviations and Differences Page 78 of 87 SYSTEM MALFUNCTION SA2 Initiating Condition -- ALERT Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful.
Operating Mode Applicability: Power Operation Startup Hot Standby Example Emergency Action Level:
- 1. Indication(s) exist that indicate that reactor protection system setpoint was exceeded and automatic scram did not occur, and a successful manual scram occurred.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 79 of 87 SYSTEM MALFUNCTION SA4 Initiating Condition -- ALERT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT TRANSIENT in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with safety systems for greater than 15 minutes.
AND Either of the following: (a or b)
- a. A SIGNIFICANT TRANSIENT is in progress.
- b. Compensatory non-alarming indications are unavailable.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 80 of 87 SYSTEM MALFUNCTION SA5 Initiating Condition -- ALERT AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. AC power capability to site-specific essential busses reduced to a single power source for greater than 15 minutes AND Any additional single failure will result in station blackout.
Differences:
- 2. Initiating Condition of SA1 in GGNSs EALs was reworded to use Div. I or II ESF instead of essential.
- 3. EAL #1 was reworded to use 15AA or 16AB instead of essential busses.
This is to clarify the intent of the EAL with respect to the power distribution arrangement at GGNS. The intent of the EAL has been maintained.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 81 of 87 SYSTEM MALFUNCTION SS1 Initiating Condition -- SITE AREA EMERGENCY Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. Loss of power to (site-specific) transformers.
AND Failure of (site-specific) emergency generators to supply power to emergency busses.
AND Failure to restore power to at least one emergency bus within (site-specific) minutes from the time of loss of both offsite and onsite AC power.
Differences:
III ESF instead of essential.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 82 of 87 SYSTEM MALFUNCTION SS2 Initiating Condition -- SITE AREA EMERGENCY Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful.
Operating Mode Applicability: Power Operation Startup Example Emergency Action Level:
- 1. Indication(s) exist that automatic and manual scram were not successful.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 83 of 87 SYSTEM MALFUNCTION SS3 Initiating Condition -- SITE AREA EMERGENCY Loss of All Vital DC Power.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 minutes.
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 84 of 87 SYSTEM MALFUNCTION SS4 Initiating Condition -- SITE AREA EMERGENCY Complete Loss of Heat Removal Capability.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. Loss of core cooling and heat sink (PWR).
- 1. Heat Capacity Temperature Limit Curve exceeded (BWR).
Differences:
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 85 of 87 SYSTEM MALFUNCTION SS6 Initiating Condition -- SITE AREA EMERGENCY Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. a. Loss of most or all (site-specific) annunciators associated with safety systems.
AND
- b. Compensatory non-alarming indications are unavailable.
AND
- c. Indications needed to monitor (site-specific) safety functions are unavailable.
AND
- d. SIGNIFICANT TRANSIENT in progress.
Differences:
None Deviations:
None
GGNS NEI EAL Deviations and Differences Page 86 of 87 SYSTEM MALFUNCTION SG1 Initiating Condition -- GENERAL EMERGENCY Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses.
Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:
- 1. Loss of power to (site-specific) transformers.
AND Failure of (site-specific) emergency diesel generators to supply power to emergency busses.
AND Either of the following: (a or b)
- a. Restoration of at least one emergency bus within (site-specific) hours is not likely OR
- b. (Site-Specific) Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.
Differences:
III ESF instead of essential.
- 2. RPV level cannot be maintained > -192 inches is substituted as the site specific indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.
Deviations:
None
GGNS NEI EAL Deviations and Differences Page 87 of 87 SYSTEM MALFUNCTION SG2 Initiating Condition -- GENERAL EMERGENCY Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.
Operating Mode Applicability: Power Operation Startup Example Emergency Action Level:
- 1. Indications exist that automatic and manual scram were not successful.
AND Either of the following: (a or b)
- a. Indication(s) exists that the core cooling is extremely challenged.
- b. Indication(s) exists that heat removal is extremely challenged.
Differences:
Deviations:
None
Enclosure I State and Local Concurrence
State of Louisiana Department of Environmental Quality M . J. -W-, KV'FOSTER, JR . L . HALL HOULINGER, GOVERNOR SECRETARY November 20, 2003 Mr. Jerry C. Roberts, Director Nuclear Safety Assurance Entergy Operations, Inc .
Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 Sub : Review and Approval of Grand Gulf Nuclear Station Emergency Action Level Revision
Dear Mr. Roberts :
The State of Louisiana Department of Environmental. Quality has recently received a request (ref.
GEOCO : 200300095) from Mr. Fled Guyno, Manager of the Emergency Preparedness department of Grand Gulf Nuclear Station (GGNS), to review and approve the revision of emergency action level (EAL) scheme of GGINS, specifically, replacement of the existing NUREG-0654 methodology by the NEI 910L Rev . 4 methodology.
It is established that the U.S. Nuclear Regulatory Commission (NRC) has endorsed the NEI 99-01, Rev . 4 methodology as a viable alternative to the NUREG-0654 methodology to develop nuclear power plant emergency action levels (EAL) (ref. NRC Regulatory Issue Summary 2003-19, dated October 8, 2003) . The Louisiana Department of Environmental Quality (DEQ) understands that the NEI 99-01, Rev. 4 methodology focuses more toward the safety of the plant, provides clearer and more precise guidance for the plant operators in recognizing and identifying emergency action levels leading to appropriate emergency classification levels, thus enhancing the operators' rapid emergency response ability. DEQ further understands that the adoption of the NEI 99-01, Rev . 4 methodology by (hand Gulf Nuclear Station does not adversely impact the emergency classification process, emergency response dose calculation, accident assessment and off-site protective action recommendations (PAP) and measures, and also the off-site environment .
Based on the above, the State of Louisiana, Department of Environmental Quality is hereby providing concurrence to Grand Gulf Nuclear Station in adopting the NEI 99-01, Rev. 4 methodology as a replacement of the existing NTJREG-0654 methodology, While we provide the concurrence, we understand that the new EAL scheme must be acceptable to NRC prior to ementation.
OFFICE OF THE SECRETARY P.O . BOX 4301 BATON ROUGE, LOUISIANA 70821-4301 TELEPHONE (225) 219-3953 FAX (225) 219-3971 AN EQUAL OPPORTUNITY EMPLOYER
Mr. Jerry C. Roberts, Director Page 2 November 20, 2003 Should you have questions regarding our position on the subject, please do not hesitate to contact Prosanta ChoNvdhury of my staff at (225) 219-3618 or via e-mail at Prasanta.Chawdhurv(c?la.aov.
cl'y' L. Hall Bohlinger, Sc.D' Secretary
/PC FRY GGNS A
Attachment III CONCURRENCE REVIEW FORM I have reviewed the changes to the Grand Gulf Nuclear Station Emergency Action Levels as proposed to the Grand Gulf Nuclear Station Emergency Plan. My comments are as follows:
d Concur as written Comments attached Name / Date Please return completed form to:
David Townsend Emergency Preparedness Grand Gulf Nuclear Station P.O . Box 756 Port Gibson, MS 39150 Fax (601) 437-2716
4UyV4/xVZ Attachment HI CONCURRENCE REVIEW FOR .,
I have reviewed the changes to the Grand Gulf Max Stwha Emergency Action Levels as proposed to the Grand Gulf NucIm Station Emergency Plan . My comments are as follows :
Concu ~ .r'°"; as written Co-ments attached Please return completed form to-,
Grand WNuclear Station Attn : David Townsend Post Oflee Box 756 Fort Gibson, Mississippi 39150 OR Fax to David Townsend at (CS11) 4370716
STATE OF MISSISSIPPI DAVED RONALD MUSGROVE, GO-VERNOR MISSISSIPPI EMERGENCY MANAGEMENT AGENCY ROBERT R. LATRAM, JR.
DIRECTOR November 70 2003 NI F. Guqymim Emergency Preparedness Grand Gulf Nuclear Station Post Office Box 756 Port Gibson, Mississippi 39150 the REF: Review of Changes to Grand Gulf Nuclear Station Emergency Action Levels
Dear W. Guynn:
I acknowledge that I received and understand the Grand Gulf Nuclear Station information on do Emergency Action Level revision from the NUREG-0654 methodology to to JQEJ 99-O I Rev. 4 methodology. I fully concur with the Grand Gulf Nuclear Station Emergency Action Levels as written and approve the proposed changes .
Should you have any questions, please contact Bill Brown, MEMA REP Program Manager, at (601) 366-3163 or bbrmynamsema .M.
RRL:BR&b POST OFFICE BOX 4501 -JACKSON,MISSISSIPPI -39290-4501 -111-IONC601-352-9100 EMERGENCY 12001224162 (24 hK)UR)
IM 1 OW400 M2
AzxhmzA III CONCURRENCE REVIEW FORM I have reviewed the changes to K CGhraund Gulf Nuclear Station Emeigency Action Levels as proposed to the Grand Gulf bhokesar Station Emergency Plan . My connnents are as follows-,
Comments attacked Organization Please tetum completed form to; Grand Gulf Nuclear Station Attn : David Townsend Post Office fox 756 Poxt Gibson, Mississippi 39150 OR Fax to David Townsmd at (601) 437-2716
III Attachment CONCURRENCE REVIEW FORM I have reviewed the changes to the Grand Gulf Nuclear Station Emergency Action Levels as proposed to the Grand Gulf Nuclear Station Emergency Plan. My comments are as follows:
DIRECTOR Concur as written Comments attached Z//e/42-t Name / Date CLAIBORNE COUNTY CIVIL DEFENSE Title / Organization Please return completed form to:
David Townsend Emergency Preparedness Grand Gulf Nuclear Station P.O . Box 756 Port Gibson, MS 39150 Fax (601) 437-2716
Enclosure 2 Simplified Drawing of Distribution Switchyard
Grand Gulf Nuclear Station Distribution Switchyard
Enclosure 3 Basis for Radiological Effluent Initiating Conditions
Date: 9/27/2004 To: Central File (/-Y)
From: Roger C. Tolbert
Subject:
Basis for Radiolo ffluent Initiating Conditions at Grand Gulf Nucle Station GIN/2004-00620 the The purpose of Effluent Initiating Conditions(IC) and Emergency Action Levels (EALs) is to provide classification thresholds for unplanned and/or uncontrolled releases of radioactivity to the environment . The four radioactive release classification thresholds addressed by this document are:
NOUE - AU1 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 2 times the radiological technical specifications for 60 minutes or longer Alert - AA1 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times the radiological technical specifications for 15 minutes or longer Site Area - AS I Offsite dose resulting from an actual or e of gaseous radioactivity exceeds the 100 mR TEDE or 500 mR or projected duration of release General - AG I Oft'site dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for actual or projected durati of the release using actual meteorology .
AU11AA1 Basis Both ICs AUI and AA l are based on a potential or actual decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments (Radiological Effluent Technical Specification (RETS)y for an extended period of time . Grand Gulf Nuclear Station defines its radioactive effluent release regulatory commitments and methodology to meet the regulatory commitments in the Of(site Dose Calculation Manual (ODCM) . In accordance with NEI 99-01, Rev 4, NUMARCNESP-007, the Grand Gulf Nuclear Station ODCM will be used as the bases for calculating ICs and EALs for AUI and AA1 .
Basis for Gaseous Effluent Release Points In accordance with the GGNS ODCM and ?`AEI 99-01 the ICs and EALs for gaseous radioactive effluents will be based on ODCM 6.3.10 and ODC-M 6,11.4:
LCO 6&10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 6.3.10-1 shall be OPERABLE with their alarm,11rip serpoints set to ensure that the limits OfLCO611-4arenotexceeded. The alarmltrip setpoints ofthese channels shall be determined in accordance with the CTF57E DOSE CALCULATION MANUAL (ODCM) .
LCO 6.11 .4 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be:
- a. For noble gases: :5 500 mrei?Vyr to the total body and :5 3000 mremlyr to the skin, and
- b. For all iodine-131, iodine-133, tritium and all radionuclides in particulate farm with half lives greater than 8 days: :5 1500 mrernlyr to any organ.
Because the ICs/EALs for AUl and AA1 are for classifying events based on degradation in the level of safety and NOT based on magnitude of the dose or dose rate, the ICs/EALs will be based on 500 mrem/yr (dose rate). This is in accordance with NEI 99-01 .
At GGNTS the total body setpoint is calculated in accordance with section 2.1 of the ODCM. For routine release monitoring GGNS uses a combination of General Electric Noble Gas and Eberline System Level Particulate Iodine and Noble Gas [SPING] monitors. Due to the availability of multiple detectors with higher range monitoring capability, the SPING system- was used for determination of IC/EALs .
In accordance with the GGNS ODUM Section 2.1, the total body conservative setpoint is calculated as follows .,
ODCM Conservative setpoint = PF'x DTB x Rj "
PP Product Allocation Factor= AF x SF'= 011 AF Allocation factor allowing for a total of four normal effluent release points = 0.25 SF' Safety allocation factor allowing for cumulative uncertainties of measurements 0.4 DTB mrcm/yr (3 .53E-5 x 60) / ( X/Q x SPINIG5 x V x K)
X/Q 710046 SeCIM3 SPING 5 Xe- 133 Volume efficiency factor SPING Ch 5 3 .54E-08 uci/cc/cpm SPING 7 Xe-133 Volume efficiency factor SPING Ch 7 1 .46E-03 uci/cc/cpm Conversion I 3.53E-OS &t cc Conversion2 60 seconds / minute K Total body dose factor for historical mixture 1 .51E+03 mrem/year per uCi/m3 SPING Radiation Max Flow (V) ODCM Ventilation System Monitors (arm) RI , Setpoint Radwaste Bldg DI 7NI 18 48000 1163 5.8 1 E+03 Containment Bldg D17NI24 6000 930 .1 4.65E+04 Fuel Handling Bldg D17N130 31000 3$0.0 9.00E+03 Turbine Bldg 017N136 35000 372.0 1 .86E+04 SBGTA ]317N]48 4300 1297,8 6 .49E+04 SBGT-9 D17N142 4300 1297 .8 6 .49E+04 Note that the setpoints are conservative due to the fact that an allocation factor is used as well as a safety factor and a historical isotopic mixture . The effective setpoint is 10% of the ODCM limit or 50 mremlyear, According to NLEI 99-01, factors that should be considered are:
Selpoints based on actual isotopics will result in setpoints higher than the "conservative method" setpoints.
EALs should increase logically and not overlap .
- Section 3.3 of NUMARC/NESP-007 recognizes that over-classification as well as under-classification should be avoided.
The 500 mRem/yr multipliers are specified in ICs AUI and AAI only to dis between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose rate. Releases should not be prorated or averaged.
For this reason the allocation factor will be changed to 1 .0 rather than using 0.25, The Safety Factor of 0.4 will be retained. The resulting EAL thresholds are shown in Table 1.
AU I = ODCM Setpoint CPI x (2 X ODCM / Actual ODCM Setpoint Limit) x Safety Factor. For example :
Radwaste Bldg AU l = 5$14 cpm x (2 x 500 mRem/yr) / (50 mRem/yT)
- 0.4 AA 1 was calculated in the same way except me AA I recommended limit, 200 X 500 mRem/yr, was used. The EATs for AA1 resulted in over range readings on Channel 5, therefore, Channel 7 will be used for me AA l EALs. The CPM value was derived by ratioing the detector Xe- 133 efficiencies .
ODCM Setpoint
= 50 mR/yr AU 1 AA l AA 1 Channel 5 Channel 5 Channel 7 SPING SPING Gaseous Release ODCM Limit 1000 niR/yr I ME5 tnR/yr 1 .00135 nLR/yr Channel 5 Channel 7 Point 500 jrnaKar Point ID Point ID CPM CPM CPM CPM Radwasm NNH8 DI 7N 119 5 .81E+03 4 .65E+04 4-65E+06 LUE+02 Containment Bldg WN124 D17NI25 4 .65E+04 33213+05 332E+07 MUM Fuel Handling Bldg D174130 D17NI31 9 .OOE+03 7 .2013+04 7.20E+06 1 .7513402 Turbine Bldg DON136 D17NI37 1 .8613+04 1 .49E+05 1 .49E+07 3 .61 E+02 SBODA 17NI48 D17N149 6 .49E+04 53913+05 5 .19E+{}7 1 .26E+03 I
1 su(; B D17N142 DI F+07 11613+03 AS1/AG1 Basis In accordance with NE199-01, AS I IC/EAL addresses radioactive releases that result in doses at or beyond the site boundary that may exceed 10% of the EPA Protective Action Guides (PAGs) .
AG I IC/EAL addresses radioactive releases that result in doses at or beyond the site boundary that may exceed the EPA PAGs (1000 mR ). Releases of this magnitude typically are the result of failure of plant systems which are identified by other ICs, this IC pertains to radioactive release events that may not be classified by system failure ICs . It should be rioted that EPA PAGs are based on Total Effective Dose Equivalent (TEDE) as opposed to Total Body Dose used by the ODCM, therefore, ODCM calculations cannot be used for AS 1 and AG I because the dose factors are different. N .E199-01 suggests a method for calculating AS 1 and AGI :
The monitor reading EALs should be determined using a dose assessment method that back calculates, from the dose values specified in the IC. The meteorology and source term used should be the same as those usedfor determining the monitor reading EALWROALRWAQ . This protocol will maintain intervals between the lTtLsjbr thefour classifications .
The Emergency Planning "Dosecalc Program" was used with the following parameters in accordance with NEI 99-01 . Meteorological data was adjusted to result in a dispersion constant of 7.1E-6 sec/m' which was used for the AUI and AA 1 calculations. Also, the ODCM isotopic mix used for calculating AUI and AAI was entered . Maximum building vent flows were also entered.
or count rates were adjusted and calculations performed on a single vent until the Site Total Effective Dose Equivalent for one hour was 100 mRem . One hour was selected based on NEI 99-01, page 5-A-13, "it is suggested that a release duration of one hour be assumed".
The results of the back calculation is listed in Table 11.
Radiation monitor readings were also back calculated for a "Core Damage Mix", because the isotopic mix has a higher abundance of iodine the 500 mRcm Commited Effective Dose Equivalent (CEDE) would be reached prior to the 100 mRcm TEDE. The radiation monitor count rate calculated is given in Table III. Note that the count rate for the core damage isotopic mix is much lower than the ODCM historical source term. Use of the ODCM isotopic mixture does not contain iodine which results in a much lower dose, therefore, a much higher EAL . NEI 99-01 recognizes this discrepancy . Section A.7 discusses the impact of source term (isotopic mix). While the ODCM isotopic mix is very un-conservative, the core damage mix is overly conservative due to the o noble gas ratio .
Section A.7 of NEI 99-01 also states that, "Far AS1 and AGI, the bases suggests the use ofthe same source terms usedfar establishing monitor EAL thresholds for AU1 and AA1. This guidance vided to avoid potential overlaps between effluent monitor EALs for AA 1 and ,4SL Other source terms may be appropriate, " Also NEI 9901, AS 1 basis section (page 5-A-12) states, "this Xprovides appropriate diversity and addresses events which may not he classified on the basis of plant status alone, e.g., fuel handling accident in spent fuel building" While consistency is needed between the four IC's to avoid overlapping EALs, a link from the ODCM methodology to the real time Emergency Planning methodology should be established.
NEI 99-01 states that, "emergency implementing procedures should call far a timely performance of dose assessments using actual meteorology and release information .. . .. . .. . . . dose assessment results override the monitor EALs. " At the AS I EAL, the Emergency Planning dose assessment model will be used, if available, for the emergency classification, therefore, a link should exist between the ODCM based EALs (AUI and AA 1) and the emergency dose assessment EALs {AS l and AG I)
The control rod drop isotopic mix was selected because :
It is the most conservative of the of the Emergency Plan mixes that may not be classified on the basis of plant status It is more conservative than the ODCM historical isotopic mixture used for AUI and AAL 0 ontinantly noble gas, which is consistent with the ODCM, historical isotopic re used for AU I and AA 1 .
The Emergency Planning, "DoseCalc", program was used to determine the radiation monitor countrate EALs that corresponds to 100 mRcm TEDE at the site boundary for a control rod drop mix . The radiation monitor count rate EALs are listed in Table IV. The same meteorological data used to calculate AUl and AAI was used to calculate AS 1 and AG 1 . AGI was determined by multiplying the AS 1 EALs by 10.
TABLE 11 Historical Source Term ODC14 Gaseous Release Max Flow ASI ASI AGI AGI Point (V) (clm) SPING CH7 AX-M CH3 SPING CH7 AXM CH3 Radwaste Bldg 48000 4.30E+04 2 .92E+03 4.30E+05 2.92E+04 Containment Bldg 6000 3 .44F,+05 2.34E+04 3.44E+06 2.34E+05 Fuel Handling Bldg 31000 165E+04 4.53E+03 6.65E+05 4.53E+04 Turbine Eqdg 15000 13713+05 93613+03 137E+06 136E+04 SBGT-A 4300 430E+05 326E+04 &80E+06 3 .26E+05 SBGT-B I 4300 ~ 180E+05 I 326E+04 I t8OE+06 '
H1 (Ire Damage Source Term Gaseous Release Max Flow ASI ASI AGI AG1 Paint (V) (cfm) SPING CH7 A)(M CH3 SPING CH7 Awl CH3 Radwaste Bld~ 48000 9.11E+01 620E+00 9.11E+02 6.20E+01 Bldg Containment 6000 719E+02 40611+01 719E+03 196E+02 Fuel Handling Bldg 31000 141E+02 9VOE+00 lAlE+03 460E+01 Turbine Bldg 15000 2.92E+02 1 .98E+01 2,92E+03 L9813+02 SBGT-A SBGT-B 4300 000 11211+01 1 .02E+03 1 60211+01 6.92E+01 112E+04 1 .02E+04 692E+02 6.92E+02 TABLE IV Control Rod Drop Source Term Gaseous Release Max Flow ASI AS] AGI AGI Point (V) (cfin) SPING CH7 1004 CH3 SPING CH7 AXM CH3 Radwaste Bldg 48000 1 .29E+04 8.81E+02 1 .29E+05 8.81 E+03 Containment Bldg 6000 114E+05 705E+03 1 .04E+06 7.05E+04 Fuel Handling Bldg 31000 .00E+04 2 1 .36E+03 10013+05 1 .36E+04 Turbine Bldg 15000 4.14E+04 2 .82E+03 4 .1413+05 2,82E+04 SBGT-A 4300 1 A4E+05 183E+03 1 .44E+06 483E+G4 SBGT-B 4300 I A4E+05 WE+W
.83E+03 1,44E+06 9.83E+04 The resulting EAL thresholds are listed in Table V. Channels were selected to obtain an on-scale reading for the EALs. Due to differences in monitor efficiencies, the count rate change do not correspond directly to the factor of 100 or 10 for differences between AAl and AJJ1 or AS I and AG1, respectively.
Also, SBGT count rates listed in Table V will actually result in doses that are approximately 10%
less than the 100 or 1000 mRem EAL. This is because SBGT is filtered and the calculation reduces the amount of iodine in the mixture .
Alternative EALs ive to the individual release point thresholds in Table V should be considered . This alternative would consist of developing new computer points for SPING Channel 7 and AXM Channel 3 . A computer point for SPING Channel 5 already exists . Using the existing SPING Channel 5 computer point as an example ;
Computer point C93CA030 - Turbine Building Release Rate in Ci/sec, is based on D 1714136 (SPING Channel 5 count rate, cpm) and DI 7N200D (Turbine Building Release Flow Rate, cfm),
The release rate is calculated by the following formula :
- 3.54E-8 uCi/cc/cpm
- 28317 cc/cf
- 4.0167 min/sec
- l E-6 Ci/uCi = Ci/sec Each release point already has this calculation performed using computer points C93CA030 thru C93CA035. These points could be added together to get a total Ci/sec readout .
The EAL threshold can be calculated using the monitor efficiency, EAL threshold and ventilation flow rate. For example:
SPRIG Channel 5 efficiency= 3 .54E-8 uCi/cc/epm AU1 Turbine Building Threshold counts per minute (Table 1) = 1 .49E+05 cpm Maximum Design Stack Flow Rate = 15,000 cfin EAL threshold Ci/sec =
I A9135 cpm
- 3 .54E-8 uCi/ec/cpm = 5.2713-3 uCilcc 5.27E-3 uCilcc
- 28317 cc/ef = 1 .49E2 uCi/cf 1 .49E2 uClkf
- 15000 cf/min * .0167 min/sec = 3.73E4 uCi/sec 333E4 uO/sec
- 1&6 Ci/uCi = 3.73E-2 Ci/sec Based on the isotopic infix Used for AUI the release rate that would result in a release 2 times the ODCM limit would be 3.73E-2 Ci/sec . If the Sum of all release points Ci/sec is greater than 3,73E-2 Ci/sec then AUl is reached. All the Initiating conditions could be calculated the same way using new computer points. The resulting EALs are listed in Table VI.
Liquid Effluent Monitor The AU l EAL associated with the liquid effluent radiation monitor is a valid reading that exceeds 2 X the alarm setpoint established by the discharge permit for _> 60 minutes . AA 1 EAL associated with the liquid effluent radiation monitor is a valid reading that exceeds 200 X the alarm setpoint established by the discharge permit for 2: 15 minutes .
Verification of Monitor Readings Because the EALs associated with liquid and gaseous effluent releases have some conservatism built in it is extremely important that every effort should be made to confirm the reading using real isotopic mixtures and procedure calculations . For AUI and AAI the calculations should be performed immediately following an alarm that indicates that the ODCM limit may be exceeded .
These alarms are based on a conservative limit of 50 mRem/year for gaseous releases and a factor of 2 times lower than the ODCM limit for liquid effluents . ODCM calculations are performed by computer software and will include the latest valid isotopic mix . Also, for AS I and AG 1, calculations should be performed immediately using real time meteorological data to confirm the classification .
However, if the calculations cannot be performed in the time required by me EAL don me classification should be made based on the monitor reading.
V Abnormal Rad Level Radioactive Effluent EAL Threshold AU I A'11 AS I X-011 Gaseous Release I I NG S P IDNKG SPING I
?
Point Channel 5 CPM Jor M Channel 7 CPM XOnitor ID GPlvl Monitor ID Monitor ID Radwaste Bldg D17NI18 4,65E+04 T) - -F , 1 19 N L13E+02 DON 1 19 119E+04 0174121 1 9k]F+Oi Containment Bldg DIX124 172E+05 1) 17-N 1 2,t D17NI25 1 .()4E+05 917NI27 7-05F-01 4'4 I Fuel Handling Bldg D17NI30 720E+04 AT IM A 0 5 0 02 ID 17N 131 2.OOE+04 17 ', 13' I !N' 1 .:,'t" T-~, +-i,) , 4 T
urbine Bldg D17NI36 1,49E+05 A)17 1 3",7 3 . 6 ,+02 I I D17NI37 4.14E+5 D 13 N I ~9 SBGT-A D17N148 19E+05 DQN 141) 116003 D17NI49 IAE+05 LAMY! I 9TS&O4 SBGT-B r-D17NI42 1 5 .19E+05 1 D1 7M. 4- 3, l,26E+ODON143 [14-,E-~-05 D17NI46 L9 .93L-,+04 ABLE VI Optional Abnormal Rad Level Radioactive Effluent EAL Threshold KJI 7~1 AS I I S r"" (i SPING Gaseous Release SPING Chmmel Channel Point Channel 5 Ci/sec 7 Ci/sec 7 ci/sec Monitor 11) ci/seC Monitor M Monitor, Monitor ID Total of al -w Point 3 .73E-02 New Poult 313000 3.37E+01* 7 E+02*
Points ID Point r)] Paint oint ID ed on results from ODCM calculation . Monitor efficiency used is 1 .46E-3 uCi/cc/cpm.
Based on results from DoseCalc . Monitor efficiency used in DoseCalc is 1.15E-4 uCi/ccIcpm. Efficiency is based on the EPA Wash 1400 mixture, This should be used in new computer point calculation so that consistency is maintained between the computer point calculated release rate and DoseCalc release rate .
Based on results of AS I X 10 (1000 mRcm/1 00 mRem. ). Monitor efficiency used in DoscCalc is 1.69E-3 uCi/cc/cpm. Efficiency based on be EPA Wash 1400 mixture. This should be used in new computer point calculations so that consistency is maintained between the computer point calculated release rate and DoseCalc release rate.
ATTACHMENT I DOSECALC RESULTS
Serial# 1093374016 Pale I of 4 GRAM GULF XMLZAR STATION MaROUNCY Do= PA0,7ECTION FOR 08/24/2004 14-00 oat* for Emergency 110tification Form Line Nvzu-,er Item 4ate Rx Shutdown NCT r .A .
10 A, 'Wind Difection From 0 d**reea at 2 . 119 mph 8 B, kffeczed SertQ--s if j K 8 C. stability Class Release info Duration I hra Started 06/2412004 12 :00 Type of Release Radioactive Gases 12 A . Rdltaze Rate Noble Gases 3,37E+01 Ci/Sec 12 3 . Release Rate lcdirieu 1 .699-01 ciisec 12 A . lose ream ttavent I hw 12 B . TEDTK Doze ?EDE Dose in vdkem 100 .00 217 E3 .99 OAO 002 OM 12 C . CDE Dose Thyroid Dose in mRem SR 240 .10 2 Miles 9,50 5 viles 4 .24 10 Mil,es 2 .32 15 Miles 1 .63 20 Miles 1 .26 See FOJXO'eiru7 Pages for Detailed Data on ProJectiar *---
Serial# 1033374016 Page 2 of 4 KCCIDMT TYKE : COVMOL ROD IMOP RE,5ULTS Distance TZVZ Thyroid akin 100 .CID03128 2ml*15015 26,052940 10 tides 0 .4047303 2 .3183699 0 .1886293 LS M:llej3 0 .2246443 1 .6261239 0 .115"43 20 miles 0-144 ,*21 1 .2618439 0 .0803323 Distance XIQ (sac/=3) Arrival Time width (Milos) 519 0 .0000071 1409 0 .2580058 2 Kites C-00Q0003 1443 0 .9B544CS 5 Tides CA000001 1547 2 .1474819 10 mile-, 0100000,01, 3735 3 .8166921 15 Titles 0 .000000i 1922 513053818 20 Tides 0 .00000z)e 2110 6 .676$408 vistamco External Tilhalation.
So 85 .870$115 "J .4865930 6,L46,7280 2 Mlles 2 .3366292 0 .2920301 0,2291109 5 pUeq 0 .7616556 0 .1210627 0 .0961249 10 miles 0,2831395 0 .0714217 0.049379B 15 miles 0.1415178 4-0499464 0.0318566 20 titles 0.081,1 024 0 .0386698 0.0235358 METEOROL OXCAL DATA wind Direction Stability Affected wind Sp*ad from Sigma That^ Cl*ss Sectors 2 .78 ZIA a deb manual stability A 11 j K MR== WORK" lrzDz COZMRSION FACTOR 1 .2
Serials !0533?4015 Page I of 4 0 talc in 3ciLD CTUT TURB BLPG Pt,1 3L Ht413LG VENT VF-NT 11M4T FLOW RAiE 6000 15000 31000 ACTIVITY ACTIVITY ACTIVITY Gs M11 .-LS 0 0 3 RIM - 6R CH S 0 0 SPING-MR CH-11 103610 41404 20034 015.-1-HR CH2 7011.5 2818 2364 Ay-M-MR CF4 0 0 0 RW ELM STAMIrf STAND371 V"IT GAS A Wo B PL34 IATE 480" 4300 410D ACTIVITY ACTIVITY ACTIVITY sz XON . -LR SPina-LA cF5 SPING-MR C" 12939 144433 144433 AXM-vy 013 98i 9830 9830 A]MMA ON 0 C011TAIMOIT D"A D21 - Y648B MOP,) 9 .98B-01 M V648C (R/11R) 9 .52B-al mum Am low) 2 u2lK635i CTMT LtAK RATE (c-1m) .1D 9
FIE-:141) WrA OF SAMPIrE :
Mile(s) frarn site regrees from 0 AIR SWRL8 TUMT (uCi/cc)
DOSE RAT:? imrfhr)
page 4 of 4 STATE ACF-';CY DOSE PRWECTION DATA SHEET DATZ 08124/2004 WIN INNUM 0 OFFICIAL DATA O.-M-TIOU ONLY - NOT OFPICrAr, DA' 1 data prepared as a 'what it , look ahead . This data not to be disseminated to other emargency facilities public 150TOPIC C41X T'fPZ (Check one)
Stea=m Cycle Care na:xage Fuel Handling off"o Liquid RadwaetQ Trod Drcw Vuer refined
- 2. PRWECTIOV PERFORMED AT TIME : 08/24/2004 14!00
- 3. REACTOR STATUS Ia . Reactor Shutdown(y!.14) N 3b . Time Reactor sbutdowm MA 3c . Tim since Reactor Shutdown(hrs) NA AnEnmy cuss S. t,;lh9D SPEED; 2 09 =,ph
- 6. VIM, IS FAM
- 7. EXPIOM;Xg DURUMON HOURS MED FOR PROJECTIC?J : I 'n rr.
IODTVE RM, 1"5F-OZ Ci/sc NOBEL GAS RFLa~ 3 .37E,01 cilcoc IK 1wmL KnEws kno :-39E+01 cilsec 11 . PILTMRZDJY,1 !-'1
Enclosure 4 Safety Analysis Parameters in the Emergency Action Limits
CALCULATION SHEET XC-01111-04006 q,a>- Date Vti2-104 Checked 1 .0 INTRODUCTION NE! 99-01 provides guidance on the development of parameters that are applied in the Emergency Action Levels. This calculation develops GGNS-specific calculations to support the EAL values for:
" a reactor coolant specific activity of 300 liCi/g dose equivalent 1-131 when 5% of the fuel cladding is damaged,
" a drywall dose rate when 5% of the fuel cladding is damaged, and
" a containment dose rate when 20% of the fuel cladding is damaged .
00 REFERENCES 2.1 Cross References 2.1 .1 NEI-99-01, Rev. 4, Methodology for Development of Emergency Action Levels.
2.1 .2 Regulatory Guide 1 .183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants~ July 2090 .
2.1 .3 Federal guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion, Second Printing 1989 .
2.1 .4 NUREG/CR-6604, RADTRAD : A Simplified Model for Removal And Case Eshnatkrn, dated April 1991 2.1 .5 GE 22A3759AE, Rev. 1, Containment and NSjQ Interface, August 1978.
2.1 .6 Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil, 1993 .
2.1 .7 SCR-2004-0149, RAPTOR 4.02, 2.1 .8 Regulatory Guide 1 .3, Rev. 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, June 1974 .
2.1 .9 Crane Technical Paper 410, Flow of Fluids through Valves, Fittings, and Pipe, 1988 .
2.1 .10 CTI Code Tower, Standard Specifications, Acceptance Test Code for Water-Cooling Towers, Cooling Tower Institute, February 1990 .
2.2 Relationships None .
J-w ut-16Y CALCULATION SHEET Sheet _Z_Cont On .
=UAL--
Calculation No . XC-01 111-04006 Rev. 0 Prepared By Date _l[UtC, Checked By 5e -S Date 3.0 GIVEN 3.1 Isotope Data The half-lives and gamma decay energies applied in this calculation is taken from FGR-12
[2 .1,6] while the core inventory is taken from NUREG/CR-6604 [2,1 .4]. The following table summarizes these parameters .
Table 3-1 Isoto e Data Gore Inventory Gamma Decay Isotope Group Hatf-Life (C!/MW) _J=ner MeVldis Kr-85 Noble Gas 1-0720E+l Yrs 2306E+02 2,OOE-03 Kr-85m Noble Gas 4 .4800E+G Krs 9 .110E+43 1TWEE-101 Kr-87 Noble Gas 7 .00001 Min 1,657E+04 7 .93E-01 1,488 Noble Gas 2 .8400E+O Hrs 2 .236E+04 1 .96E+00 Halogens 1040000 2 .5$1E+44 182ET1 1-132 Haloqens 2,3000E+O Hrs - 3,792E+04 2,28E+00 11-133 Halogens 2 .0800E+l Hrs SAVE 607M FAWas Noble Gas 1 5 .245000 Dys- 5A25004 4.60E-02 11-134 Halogens 5-2600E+1 Min 5 .930E+04 2 .63F=+00 1-135 Halo ens No ble Gas 6 .6100E+O Hrs 9.0900E+0 Hrs
&099E+04 1 .289E+04 t 1 .58E+00 2 .-
"'c"'vu 49E-01 4.0 ASSUMPTIONS All applied assumptions are reported in Section 5 of this calculation .
CALCULATION SHEET XC-Q 1111-04006 Date~--SK ZL-c_q_
~ CheckedBy 5.0 CALCULATIONS 5.1 Primary Coolant Activity Level (FC-1)
NEI-99-01 [2 .1 .1j indicates that a reactor coolant specific activity of 300 ;ACi/g dose equivalent I-131 would be a conservatively low value to represent the reactor coolant activity associated 5% fuel clad damage . This section confirms this statement.
This calculation will determine the reactor coolant specific activity based on .-
0 instantaneous iodine release from fuel rod gaps of 5% of the core, 0 core average gap fraction of 5% per Table 1 of Reg Guide 1 .183 [2 .1 .2),
- thyroid dose conversion factors from Federal Guidance Report 11 (2 .1 .3, 0 no RWCU cleanup removing the source terms or significant source term carryover in the reactor steam, standard core source term inventories from NUREGICR-6604 [2 .1,4],
0 core power level of 3898 MW,
- reactor coolant inventory of 6.8E5 pounds per Reference 2.1 .5.
As shown below, the reactor coolant specific activity is calculated to be nearly 1200 itCl/g .
Therefore, the 300 pCi/g appears to be a conservatively low value.
Core Coolant Coolant FGR-1i Thyroid Dose inventory Activity iodine Conc Dose Conversion Fraction Equivalent !-
Isotope (Cilmw) (Cil (pCJ/g) Factor (Rem/Cl) 1-131 DCF 131 (pCi1g) 1-131 2 .58E+04 2 .5E+05 82E+02 11 ME+06 1 .00E+00 8.15E+02 1-132 3,79E+04 3,7E+05 1 .2E+03 6.44E+03 6.96F-03 T14E+00 1-133 5.42E+04 5.3E+05 1 .7E+03 1 .80E+05 1,66E-01 2.85E+02 1-134 5.93&04 5.8E+05 1 .9E+03 1 .07E+03 9,87E-04 1 .85&00 1-135 5,10E+04 5 .0&05 I-6E+03 3.13E+04 2.90E-02 4.67E+01 1 .16E
MEntergy CALCULATION SHEET Sheet 4 Cant on -5 Calculation No, - XC-Q1111-04006 Rev, 0 5.2 Drywall Radiation Monitoring (FC-3)
NEI-99-01 (2 .1 .11 indicates that the drywall dose rate should be based on the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventories associated with the calculated clad damage used in EALY . Therefore, this section will determine the drywall dose rate associated with 5% fuel clad damage . Cases are evaluated based on source terms residing solely in the drywall and mixed homogeneously within the containment and d This calculation will determine the drywall dose rate based on:
" instantaneous iodine release from fuel rod gaps of 5% of the core,
" core average gap fraction of 5% for halogens and nobles gases per Table 1 of Rag Guide 1 .183 [2 .1-21,
" standard core source term inventories and isotopes from NUREGICR-6604
[2 .1,41,
" core power level of 3898 MW,
" isotope half-lives and decay energies from FGR-12 [2 .1 .6)
" containment volume of 1 .4E6 ft 3 (2 .1 .6J,
" drywall volume of 2.7E5 ft' [2 .1,5].
Considering that 5% of the fuel is damaged and that this damaged fuel will release 5% of its source terms, the drywall atmosphere is calculated to hold 0.0026 of the total core source term inventory.
Rag Guide 1-3 (2-1 .8] gives a typical conversion factor for the gamma dose rate associated with 3,
an inifinite cloud of gamma radiation to be 0.5 Rad/sec per MeV-Ci/diS_M which is based on the total released energy being equal to the absorbed energy. This conclusion can be shown Ci/M3 by calculating the energy release of a source term concentration of I of an isotope that lb/ft3 has an energy release of I MeV per disintegration in air at 68 'F 14 .7 Asia (p = 0.0752 1203 gkc (2 .1 .9]} .
dis El0 - - - I -
mev 6,s - - 1 .6E - 6 ergs m Ci-s MeV = Rad 174933-100 -----ffP--1 .2E-3 h Rad-g cm s Although standard conditions any be applicable for the purpose of radiological calculations to exposed individuals, the post-LOCA conditions in the containment would lead to a higher air density due to the higher pressure and humidity such that a lower dose rate would be anticipated . Assuming post-LOCA conditions of 150 *F and 5 prig, the air density would be U9215-3 g/cc.
7 + 5 psia ) f 464+68°R ~ 1 .392E-3 sia ) ~460+150 0R) cm 3
CALCULATION SHEET XC-Q 1 111-04006 Saturated air at these conditions would hold ^0.11 lbs of water vapor per pound of dry air based on extrapolation in Reference 2.1,10, The density would therefore become 1,55E-3 g/cc (111 1,392E-3). The corresponding dose rate would become 0,38 Rad/s.
For a semi infinite cloud, the dose rate would be half of this value (consistent with Reg Guide 1 .3) or 0,19 Rad/sec per MeV-Cildis-m', which can be converted to 0.0185 Rad1hr per Me'V/s-cc .
Rod s Rod 3600 0.19 --~- - hr hr
-p .01[35 Mev-a A~ / 3 AM
--'- 1740 --
dis - m ci-sbo0cm) S C A typical semi-infinite to finite cloud conversion factor for gamma radiation is vo'"" 11173, where V is the volume of the compartment in W 12 .1,4], Therefore, the gamma dose rate can be described by the following formula; v OMB b= 0 .0185-t (51) 1173 energy deposition (MeV/cc-s)
= compartment volume (ft)
Based on the RAPTOR j2 .1 .7] output in Attachment i, the dose rates were determined at several time decay times. These results are listed in Table S-1.
CALCULATION SHEET Sheet 6 cant tin -1 lculation No . XC-Q I 11 1-04006 Rev .
Prepared By 91 -~,~ . Date 5.3 Significant Radioactive Inventory in Containment (PC-4)
NEI-99-01 [2 .1 .1] indicates that the containment dose rate should be based on 20% fuel clad damage . Since no specific guidance is provided regarding the nucride groups or release timing, this section will assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventories consistent with the drywe)l dose calculation, This calculation will determine the containment dose rate based on:
" instantaneous iodine release from fuel rod gaps of 20% of the core,
" cue average gap fraction of 5% for halogens and nobles gases per Table I of Reg Guide 1 .183 [21 .21,
" standard core source term inventories and isotopes from NUREG/CR-6604
[2 .1 .4],
" core power level of 3898 MW,
" isotope half-lives and decay energies from FGR-1 2 [2 .1 .6]
" containment volume of 1 .4E6 ft' 12 .1,51,
" drywell volume of 2.7E5 ft3 [2 .1 .5] .
Since an unscrubbed source term must enter the drywell before release into the containment, the total volume of the drywell and containment will be applied to determine the source term concentration . Considering that 20% of the fuel is damaged and that this damaged fuel will release 5% of its source terms, this volume is calculated to hold 0 .01 of the total care source term inventory .
Based on the RAPTOR output in Attachment I and Equation 5-1, the dose rates in the containment were determined at several decay times . These results are listed in Table 6-1 .
CALCULATION SHEET
-)(C-Q1111-04#706 Calculation No.
6.0 CONCLUSION
S This calculation demonstrates a reactor coolant specific activity of 300 IlCilg dose equivalent I-131 is a conservative underestimate of that expected when 5% of the fuel cladding is damaged.
In addition, the following table reports the drywall and containment dose rates predicted for 5%
and 20% cladding damage.
to 6-1 Dose Rates in I and Containment
_.Energy Deposition (Mlev(cc-s). Does Rate F2adlhr _I I
Applied Drywall D ell & Drywall Drywall Drywell DrTyweR &
Volume On I Clowintlasin1jetrit
- !a-- l I Containment Containment 1 tairim t Clawd 5% 6% 20% 5% 54/0 2W]a Failure DoSe Drywall Drywall Containment Drywall Drywall Containment Rate itt:
Time ----
2007 127006 1 .31E+07 21,874 ___ 3,537 24,673 511 E+06 8.43E+05 3.37E+06 _5,634 9111 6,355 0107 12 2,96E+C)6 4.79E+05 1,92E+06 3,203 -_ . 981 51 1031
I 1 ------- 24 YnEy"I 1.01E+06 1i -- 9,694 --------- 274 L- 1,910
Calculation XC-Q1 111-04006, Rev . 0 Attachment 1, Rev. 0 Page 146 R,RS-RRR AAkk PPPPP9 TTTTTTTTTT 000 RRRRRR w"Mma AWAAMAAA PPPPPPPP TTTTTTTTTT 0000000 RRRRRRRR Rk RR AA AA PP LIP TT 00 00 RR RR RR RR AAA AAA PP PP TT 00 00 RR RR.
RR RR AA AA Pp PP IN 00 On RR RR RRRRRRRR AAAAAAAAAA PPPPPPPP TT 00 4 .02 00 RRPRRRRR R"RUR ARAVASAA PPPPP? TT 00 00 wwRRRRR RR RR AA AA PP IN 00 00 RR RR RR RR AA AA VP IT 00 00 RR RR RR RR AA AA PP IN 0000000 RR RR RR RR AA AA PP IN OCID RR RR 111iiTTill NW NN PPPPPP Uly VU TTTTTTTTTT T111111111 XNN NO PPPPPPPP UU U'V TTTTTTTTTT 11 NNNN ON Pp PP UU UU IN II N14 NN NN Pp Pp UU VU IN IT NN NN NO PP PP 1313 UU TIT IT RN NN NN PPPPPPPP UU UU IN 11 NO NN NN PPPPPP UU UU IN IT NN NN NN PP UU UU IN 11 NN NO NN PP MYU UUU TT IIIIIIIIII KN NNNV Pp UUUUMUU TT 111111ilTr ON NN Pp UU 'PT Execution Time : 13 .36 :42 on 08/11/04 CASB DESCRIPTION Cal(- Ulation XC-QIIII-04006, Rev . 0 1 Evaluation of Emergency Action Levels Generation the Drywell and CMT energy deposition rates based on ;
5% C-f clad into the drywall - 0 .25V of the core inventory 21 0V ol clad into the containv.,, ent - 1% of the Core inventory MODELED NUCLIDE PARAMETERS Standard GGNS LOCH Isotopes Alpha Beta Gamma Isotope Group Half-Life (Mev! (Mev) (Mev) ri-85 Noble pas 1 .0'720P .001 Yrs 0 .000DEvOOO 2 .5100E-001 2,0000E-003 KY , b5,1n Noble Gas 4 .46006*000 Hr S 2 .5500R-00a 1,580OR-001 JKj-S? Noble Gas 7 .6300E+001 Min O .OOOOE+OOU 1 .3240E+000 7 .9300E-001 Ki-8B Noble Gas 2-8400E+000 Hrs 0 .0000E+000 3 .64GOE-001 1,9550E+000 1 , 131 Halogens 8,0400E+aOO Dys O .OOOOE+000 I .R200E-001 3,8200E-G01 1 " 132 Halogens 2 .3000E+000 Hrs 0,0000S+000 4 .95OOE-001 2,2800E+000 1 133 Halogens 2 .0900E+001 Hrs 0 .0000E+000 4 .110CE-D01 6 .0700E-001 Xa 133 Noble Gas 5 .2450E*D00 Dys 0, 000C1?+UO 4 .6ZOOZ-002 T - 134 Halogens 5 .2600E+001 Min 0-0000F+000 2 .625tE+000 S-135 Halogens 6 .6100E , 000 Hrs 0 .00DOE-000 3 .67OOE-001. 1,57609+000 Xa135 Noble Gas 9 .0900E,000 HrS 0-0000E+000 3,17002-001 2,49OOE-001
Calculation XC" 01111-04006, Rev. 0 Attachment i, Rev. 0 Page 2 of 6 IKODI L PAP' AMSTETzs Care Power Level = 3,8960E+003 MW Decay Time DDDQFi, 0QD See in the core .
Calculation End Time -- 2,4000E+001 lfrs NODE PAP.AMETERS Dxywell (5v)
Volume -~, 2,7000E+005 cu .ft .
Decay Enabled, Daughter Product Tzatking Not Enabled .
Inventory Printed in Cozies .
Energy Deposition Summarized Each Timestep .
Integrated Activity Not Tracked .
DW & CMT (5k)
Volume = 1 .6700E+006 cu,ft .
Lecay Enabled, Daughter Product Tracking Not Enabled .
Inventory Printed in Curies .
Each Energy Deposition Summarized Timeatep .
integrated Activity Not Tracked .
BW I CMT (2040 VVolume - 1 .6100E+OD6 cu .ft .
Eecay Enabled, Daughter Product Tracking Not Enabled .
ntory Printed in Curies .
rgy Deposi.tjon Summarized Each Timestep .
egrated Activity Not Tracked, RFLIASE POINT(S)
There are no release points .
RECEIPT POINT(S)
Theie are no receipt points .
INITIAL INVENTORY MI-85 in Drywell 15%) at 2,5002-003 of 2,506GE+002 Ci/NIW Kr-BS In DW & CMT 15t) at 2 .SODE-003 of 2-5060E+002 Ci/MW Kr-BS In 1)V7 & CMT (20V at I .ODOE-002 of 2 .5060E+002 Ci/MW Kr~85m in Drywell (5%) at 2-50CE-003 of 9 .1100E+003 C11MW Xr 85M 171 DW & CMT (5k) at 2,500TI-003 of 9 .3100R+003 Ci/MW Kr 85m In DW & (>VT (20% at 1 .000E-0GZ of 9 .11OOE4003 Ci/MW Kr-87 In Drywall f5W) at 2 .500E-003 of 1 .6570E,004 Cj,IMW K.-Q7 In DW & CMT t5%) at 2 .500F-003 of 1AS70E+004 Ci/MW KI-87 In DW & CMT (204 at 1 .000E-002 of 1 .6570E+004 Ci/MW Y,:-89 In Drywall (51) at 2 .500E-003 of 2 .2360E+004 CJ/MW Xi-88 In DVI & CMT (5%) at 2 .50CE-003 of 2 .2360E+004 Ci/MW K1 -BB In Dl?i F. C14T MV at of 2,736Q,E-V9i WMW 1 131 In Orywell (5%) at 2 .50CF-001 of 2,5AI0E+0Qq C i ( 1,414 L002+000 elem . 0 .00E+000 org . 0 .00F+000 part, I131 In DW & CMT (5%) at 2 .500S-003 of 2 .5810E+004 Ci/Kw 1,00E+000 elem . 0,00E+000 org . G .GUE+000 part .
I-131 In DW & CMT (20t at 1,000E-002 of 2 .58109+004 Ci/t4w I'00E+000 elem . 0,00F .000 erg . 0,00E+000 part, S - 1'32 In Drywall 150 at 23"ODO of 3,MMM CaKW noMon slam . VUWO" oa . 0 MM" Part,
Calculation XC-Q1 111-04006, Rev, 0 Attachment 1, Rev . 0 Page 3 W 1-13a in DW & C"T (5%1 at 2 .509E-U03 of S,7920Z+0C4 Ci/MW 1 .00Z'000 elem . 0 .00E-000 or3 . 0,00E+000 part .
1 132 In Dw & CMT {20% at 1 .000E-002 of 3 .7920E+004 Ci/MW 1 .002+000 clam . D .40E+090 org . D .ot'71 +0 00 part, 1 , 03 Ln Drywall WO at 2 .50DE-003 of 5 .4170E+004 Ci/MW I .OOE+000 elem . 0'00E+000 org . O,OOE+000 part .
1-133 In DW & rMT SSA) at 2 .SDDE-003 of 5,4170E+004 Ci/t4w 1 .00E+000 slam . Q'00B+000 019, 0 .00Z+000 part, 1 - 133 In PW & CMT (201 at 1 .000E - 002 of 5 .4170E+004 Cl/MW 1,00E+000 elem . 0 .00F+Ozo org . 0,00F4000 part .
1-13 1, Sn Drywall at 2 .500E-003 of 5-9300F'004 Ci/MW 1'00R+Doo elem . 0 .04E+000 org . 0 .00E+000 part .
1-134 In OW & CMT at 2 .5002-003 of 5 .9300E+004 Ci/MW I .ODE+000 elem . O .DOE+000 org . Q .OGE+000 part .
1-134 In QW & CMT (20% at .444E-002 1 of 5 .9100e+004 ci/Mw I .ZOF .004 elem . D'on+000 org, O .GOE000 part .
1 - 135 1n Drywall at 2 .500E-003 of 5 .0990E+004 Ci/MW ! .00Fi000 .
elem 0,04E+000 org . 0 .002+000 part .
1-135 In 1)w & CMT at 2 .500E-003 of 5 .09909+004 Ci/MW 1 .00E+000 elem . 0 .00P+000 org . .04E+000 4 part, 1-135 In DW S, CMT f2()V at 1 .000E-002 of 5 .0990E+004 Ci/t4W 1 .00P+000 elem . 0 .00E+000 org . O .OOE+000 part .
XP-133 Zn Drywall (5%) at 1 .500E-003 of 5-4250?+004 Ci/MW xe-133 in Dr.? & CMT (S k) at 2 .500E-003 of 5 .4250E+004 Ci/MW of Xe-131 In DW & CMT (20% at 1 .000E - 0II2 5'4250B+004 C4/Mw Xc - 135 In -Drywell (5 t) z't 2'500E-003 of 1 .2894F4004 Ci/MW xe-135 in DW & CMT at 2 .50GE-003 of .2a943 1
- +004 Ci/MW Xe-135 In VW & CMT (20% at 1,000F-002 of 1 .2890E+004 M/MW ULFASE PARAMMMS There are no releases .
FLOV PARAMETERS flay #1 from Drywall (5%) to DW & CMT (51)
G"On"o Sirs to 1-2"U"i Hys at 00"M"" Am FlUER PARAMETERS These a" no &In" .
RRM(VAL PAR7LM3TERS Theie are, no removal mechanivms, DIFIUSION PARAMETERS Theca are no difftsion parameters .
DOSI LOCATIONS These are no dose locations .
Calculation XC-Q 1111-04006, Rev . 0 Attachment 1 . Rev . 0 Page 4of6
?-P,F,RRP AAAA PpPppP TTTTTTTTTT 000 PRRiZRR RKRRRRR PXMw PPPPPPPP TTTTTTTTTT 0000000 RRRRRfMR FR PIP, AF . AA P? PP TT 00 00 RR RR AAA RR RR AAA PP PP TT 00 00 RR Rp, RR PR AA AA PP PP TT 00 00 RR RR RRFRRBRR A-AAAAAhAAA PPPPPPPP TT 00 4 .02 00 RRRRRRRR 00PROR AAAAAAAAAA PPPPPP 71 00 QQ W"RaRRURR AA PP K 00 00 RR R-4 RR, RR AA RR RR AA AA, PP TT 00 00 RR RR RR RR AA AA PP TT 0000000 RR RR RR RR AA AA P? TT 000 RR RR (1 00 Tj UU 7TTTTTTTTT PPPPPP UU UU TTTTTTrTTT 0000000 UV TTU TTTTTTTTTT PPPPPPPR 'JU [ICS TTTTrTTTTT OC 00 UU LIU TT PP PP UU UU TT (110 00 In LTU TT Pp Pp UU UU TT 00 00 UTJ UU TT Pp Pp UU UU TT 00 00 UT) UU TT pPPPpPPP UU UU TT 00 00 UTJ UL TT pppppp Ulu UU TT 00 DD W UU TT FP VU uu TT Dr 00 VVU UVU TT PP UUU UUU TT (OGN&O uVrJAIX&TVw TT PP UUU=TLrj 'ST coo UU TT PP UU TT 0 000n"o sec TORY DISTRIBUTION Is< tope DryWell (5t) DW & CMT (5t)
Curies C-uriev XI-85 2 .442097Z+n03 2-,42091E+Q0) 9-168388R+003 Kr-85m 8 .871695E+004 8 .977695E+004 3 .551078E+005 KI-87 1 .63.4746E+005 1 .614746E+005 6,4589869+005 K1 , 88 2 .178982E+005 2,178982E+005 8,715928E+005 1-131E 2 .5151B4E+001, 2 .515184E+005 1 .006074E+006 I -1 32E 3 .695304E+005 3 .695304E+005 1,478122E+006 1-133E 5,2788EE9+005 S,278S66r+005 2,L11547E+006 Xe-133 5,286662'3+005 5 .286662E+005 2 .114665E+006 1 . 134F 5 .778785E+005 5 .',178785E+005 2 .311514F,1+006 7 135E 4 .968975F+005 4,968 .975E005 1 .987590E+006 Xe 1.35 i . 251 3-2561302+035 5 .0245222+005 F,NEfGY DEPOSITION Rad :atiOn Type ALPRA ALPHA BFTA BETA GAMMA 0""1 (Mev/s/ccl (Mev/ccl, imev/s/cc) (Mev/cc) (Mev/s/cc) (Mev/cc)
Irywell (5s) 0 .000000E+000 OA00000E+000 6 .862530E+006 0,0000009+000 2,024206E+007 0 .000000E+000 D4 & CMT (5*) 0-000000E+000 G,0Q0000E+0V0 1 .109511F+006 0 .000000E+000 3,272648F,+006 0 .000000R+000 Di & CMT (20% 0 .000000t+000 4,438043E+006 0 .000000E+000 1,309067F+007 0 .000000E+000
Calculation XC-Q1 111-04006, Rev, 0 Attachment 1, Rev . 0 Page W6 Time - .0 .004E+ Hrs 6
TNVFNTORY DTSTRIBUTION Isotope Drywell (St) DW & CMT (SV) DW & CMT (20V Curies curies curies KY-SS 2 .44I9S9S+00'3 5 .767956R+003 Kr-65m 3 .50863.OE+004 3 .iO8610E+004 1 .403444E+005 KI-87 6,134503E+003 6 .1345039+003 2 .453801E .004 Ki-89 5 .038192F+004 S .0382929+004 2 .015277E4005 1-131E 2 .461554E+005 2,461554E4005 9 .846218E+005 I-112'Z 6,QS8352E+004 6 .05aI52S+004 2 .4211SIZ+t}05 1-133E 4 .322202E+005 4 .322202E+005 1 .728881FOO6 Xe-133 5 .114853E+005 5,114953E+005 2 .045941E+006 1-1348 5 .029851E+003 5,029851E,003 2 .011940E+004 1-115E 2 .64B6e5R-OOS 2 .64861)SE+005 1 .0594422+006 Ye-135 7,S49413F,004 7 .949413V.+004 3 .179765E+OOS ENERGY DEPOSITION Rad :Ation. Type ALPHA ALPHA BETA BETA GAMMA GAMMA.
(Mev/s/'Zc) (MeVICCI (Mev/s/ccl (Mev/ccl Nev/slccV (Mev/ccl 0 .00000DE+000 0 .0000ODn4000 2 .351997S+006 2,0$29922+011 D~ & C14T (5%) O .OOOOOOZ+000 0 .00000')F,+000 3,802629E+005 1 .287833E+010 8 .429607E00S 3,36'7712E+010 D; & CMf 120$ 0,00000OF4000 0,0000006+POO 1 .5210529+006 S,151330E+010 3,3719438+006 1,347085E+Oll Time = 1. .100E+001 lira INVINTORY DISTRIBUTION IS(tope Dr -ywell (SW) DW & CMT (5t) DW & CMT (20V Curies Curies Curies K-65 2 .411881E+00 2,4418$IE+003 9 .767523E+003 Kr , ASm 1 . ~86660R+004 5 .546638E+004 Ki-87 2 .330529E+002 2 .330529E4002 9,322114E+002 K)-88 1,164919r+004 1 .164919E .004 4 .659677F+004 1-131F P .409068E+ooS 2 .4090C,8E+005 9,636272P+00!,
1-132E 9 .9325068+003 9 .932506E+003 3 .973003E+004 1 . 133E 3 .538910Ef005 ?,538910E+005 1 .4155643+006 Xe 133 4 .948627E+005 4 .948627E+005 1 .979451r-,006 1-134E 4 .377580E+001 4 .3779809+001 1 .751192E+002 T-135E 1 .411782E+005 1,411782EO05 5 .647128S4005 Xe-135 5,030781E+004 5 .030781?+004 2,012312E+005 ENBkGY DZPOSXTJOX Radiation Type ALPHA. ALPHA BETA BETA GAMMA GAMMA tl'lewa/cc) (Mev/cc) mpv/a/cc) (Me'q/cc) (Mpv/s/cc) (Mev/cc)
CalculaVion XC-0 I 111-04006, Rev. 0 Attachment 1, Rev. 0 Page 6&6 Iry.ell (5t) o, oooooo8 .owG o, i . 6413SSl', I 006 1,2162792 .011 2 .964343E+03S 2 .9224152.011 DV CMT '5%) U .000000S+Ooo o,ooooooE+ooo 2 .6634D9E+005 1,966439E+010 4,7g2650E+005 4,724862E,#OIO DI( CMT (201 0 .000000E+000 1 .065364E4006 '? ,86S'155E.010 1 .91'7060R+006 I .B89545E+011 Time- = 2,400E+001 Hrs INVFNTORY DISTRIBUTION Isctope Orywell (5%) DW L CMT (20%
Curies Curies Kt-A5 2,441665E+003 2,441665E+003 9 .7666596+003 xr , 85ln 2,165906P4003 2 .165906E+~G03 8 .6636223+003 Kr-87 3 .3G360IE-001 1,145441S+05t KI-68 6,227849E+002 6,227849E+002 2 .491140E+003 1-131E 2,101429L+ODS 2 .207429E+005 9 .229715E+005 1-3321K 2 .6697312+002 2,669731E+002 1,0678922+003 T 133E 2 .372457E+005 2 .3724$7S+OO5 9 .48982$E4005 Xe-113 4,632206C4005 4,632206E+005 1 .852882E+006 t-134r 3 .31673CE-003 3 .316736E-003 1 .326694E-002
.135E 1 4,011145F+00-1 4 .011145E+004 1,604458E+005 Xe-135 21 ,0148195+004 2 .014819S+004 8 .059277E+004 ENERGY DRPOgtTION Rad-atforl Type ALPHA ALPHA BETA BETA GAMMA GAMMA (Mev/s/cc) (14eviccl (Mev(S/cc) tMe'VIcc) (Mev/s/cc) (Mev/c(,.)
frywell t5w) 0,00000OF-000 0 .000000E+000 1 .100692E+006 1,7915572+011 I .S67324E+006 3,839021 .2+011 AS & CMT ;S'k) l,7l5S45E+Do5 2 .996529£+010 2 .53399"tE+005 6,206901E+410 DP & CMT (20k 0,000000E+000 0 .000000E+000 7,118181E+005 1 .013S99E+006 2 .482720V+Oll