ML20210J303
| ML20210J303 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/27/1999 |
| From: | Sen G VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0654 BVY-99-94, NUDOCS 9908040244 | |
| Download: ML20210J303 (49) | |
Text
VERMONT YANKEE
.-y NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257-5271 July 27,1999 BVY 99-94 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 i
References:
(a)
USNRC Emergency Preparedness Position (EPPOS) No. 4," Emergency Preparedness Position (EPPOS) on Emergency Plan and Implementing Procedure 4
Changes," dated November 19,1998 (b)
NUREG-0654 FEMA-REP-1, Rev.1," Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear i
Power Plants" (c)
NUMARC/NESP-007 " Methodology for Development of Emergency Action
]
Levels," dated January 1992 j
(d)
Letter, USNRC to VYNPC, "NRC Inspection No. 50-271/95-07," dated June 12, t
1995 (e)
US'NRC EPPOS No.1," Emergency Preparedness Position (EPPOS) on Acceptable Deviations from Appendix 1 of NUREG-0654 Based Upon the Stafi's Regulatory Analysis of NUMARC/NESP-007, Methodology for Development of Emergency Action Levels," dated June 1,1995 (f)
EPA-400-R-001," Manual of Protective Action Guides and Protective Actions For Nuclear Incidents," dated May 1992 (g)
NUREG-1228," Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents" (h)
NUM ARC," Questions and Answers - Methodology for Development of Emergency Action Levels, NUMARC/NESP-007 Rev. 2," dated June 1993
Subject:
Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271) f Proposed Chances to Emercency Action Levels The purpose of this letter is to submit proposed changes to our Emergency Action Levels (EALs). Based i
on the expectations established in Reference (a)_ prior NRC approval is required for non-administrative changes to EALs. Therefore, we request that the NRC review and approve the enclosed material regarding the technical changes to our EALs. These EALs will be included in the Vermont Yankee Emergency Plan and Emergency Plan Implementing Procedure AP 3125," Emergency Plan Classification and Action Level Scheme." These changes have been concurred with by the states of Vermont, Massachusetts and New Hampshire.
l'0[ provides a listing of the changes to the EALs along with a reference to the bases documents
\\
supporting the change. Attachment 2 provides a cross-reference between NUREG-0654 and our EALs. provides a listing of deviations and clarifications from the guidance provided in NUREG-0654 Appendix 1.
9900040244 990727 PDR ADOCK 05000271 F
E VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-94 / Page 2 of 2 We would appreciate your review and approval by October 15, 1999 to support our present implementation schedule. Please contact us if you have any questions or need additional information.
f l
Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION 77 h A
l i
Gautam Sen Licensing Manager I
l I
Attachments I
cc:
US'NRC Region 1 Administrator
' USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS Richard J. Conte, Chief, Emergency Preparedness and Safeguards Branch, Division of Reactor Safety, USNRC Region i Vermont Department of Public Service
Docket No. 50-271 BVY 99-94 Vermont Yankee Nuclear Power Station Summary of Emergency Action Level Changes i
I i
BVY 99-94/ Attachment 1/Pege 1
SUMMARY
OF EMERGENCY ACTION LEVEL (EAL) CHANGES:
- 1) Revised Emergency Action Levels (EALs) to reflect acceptable deviations from NUREG-0654.
Updated EALs to reflect changes in the USNRC's position on acceptable deviations from NUREG-0654, Appendix 1 as an overall enhancement to the emergency plan. Rearranged EALs to a one sheet format which includes the " Fission Product Barrier Matrix" to increase the effectiveness of the EALs when being used to classify an emergency.
- 2) Added mode applicability to each event initiating condition specific to Vermont Yankee.
Added mode applicability to each event initiating condition consistent with bases described in NUMARC/NESP-007 and specific to Vennont Yankee's EALs, to improve the ability to identify and characterize event initiating conditions during any operating mode. This item was the subject of a commitment to the NRC by Vermont Yankee to add mode applicability to each EAL. Mode applicability defaults to all modes except where plant or other conditions preclude initiation of the event from a particular mode or mode as defined in or limited by Technical Specifications. Technical bases under the example EALs in NUMARC/NESP-007 are allowed to be used to enhance and clarify site specific EALs developed from NUREG-0654 by i
USNRC EPPOS No.1.
I The identified operating modes are defined as follows:
Power Operations - Reactor is critical and the mode switch is in RUN.
Startup/ Hot Etandby - The mode switch is in STARTUP/ HOT STANDBY.
Hot Shutdown - Mode switch is in SHUTDOWN and reactor coolant temperature is >212 F.
Cold Shutdown - Mode switch in SHUTDOWN and reactor coolant temperature is s 212 F.
Refuel - Mode switch in REFUEL and reactor coolant temperature is s 212 'F.
Defueled - Reactor Pressure Vessel contains no irradiated fuel.
The plant operating mode that existed at the time that the event occurred, prior to any protective sptem or operator action initiated in response to the condition, is compared to the mode applicaoility of the EALs. If an event occurs, and a lower or higher plant operating mode is reached before the emergency classification can be made, the declaration shall be based on the mode that existed at the time the event occurred. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold Shutdown or Refueling for mode applicability, even if hot shutdown (or a higher mode) is entered during any subsequent heatup. In particular, the Fission Product Barrier Matrix EALs are applicable only to events that occur in Hot Shutdown or higher. A reactor coolant temperature greater than 212 F would be indicative of being in a mode of Hot Shutdown or higher regardless of mode switch
BVY 99-94/ Attachment 1/Page 2 position, i.e. temperature where specified outranks the switch position. Mode temperature of 212 F is based on Vermont Yankee Technical Specifications below which Technical Specifications formMg a basis for the EALs are relaxed or as in the case of shutdown cooling, become the limit.
Mode applicability is consistent with the additional bases provided in the NUMARC document except as noted in the EAL and EAL technical bases.
- 3) Revised events to reference Technical Specifications directly.
Revised events which restated Technical Specifications to directly refer to the Technical Specifications to clarify and simplify EALs. Alarms, samples or other bases were used where applicable. Entry levels were reviewed to not be more restrictive than Technical Specifications to eliminate over-conservative entry into EALs which unnecessarily activate state, local, and NRC facilities. Declaration of an emergency is generally not required if a plant remains within the approved operating envelope of an Action Statement.
Pursuant to USNRC EPPOS No. 4, the changes in EAL do not adopt a methology other than one approved by the NRC, or change a setpoint for an EAL. EALs deleted are; 1) incorporated into other related EALs, 2) allowed to be removed per USNRC EPPOS No.1,
- 3) more effectively addressed by other EALs, or 4) not referenced in any NUREG examples /NUMARC bases. EALs added comply with the deviations addressed in EPPOS Nc.
1, the Q&A for EPPOS No.1, or are intended as a transition between EALs shown as a NUREG example entry level or higher.
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Docket No. 50-271 DVY 99-94 Vermont Yankee Nuclear Power Station NUREG-0654 to AP 3125 Cross-Reference
V BVY 99-94/ Attachment 2/Page i Vermont Yankee Nuclear Power Station NUREG-0654 to AP 3125 Cross-Reference NUREG-0654 IC AP 3125 EAL Unusual Event 1
See Attachment 3*
2 U-1-a 3a U-2-b, See Attaciunent 3*
3b U-2-a 3c See Attachment 3*
4 See Attachment 3*
5 U-3-a 6
See Attachment 3*
7 U-6-a 8
See Attachment 3*
9 See Attachment 3*
10 U-4-a 11 U-7-b, U-7-c, See Attachment 3*
12 U-9-a, See Attachment 3*
13a U-5-c 13b U-5-a 13c U-5-d 13d U-5-b 14a U-8-b 14b U-8-b 14c U-8-a 14d U-8-d 14e U-8-c 15 U-10-a, U-7-a, See Attachment 3*
16 See Attachment 3*
17 See Attachment 3*
~
- - VYNPS EAL NUREG-0654 Appendix 1 Deviations & Clarifications
BVY 994 %ttachment 2/Page 2 Vermont Yankee Nuclear Power Station NUREG-0654 to AP 3125 Cross-Reference NUREG-0654 IC AP 3125 EAL Alert la A-2-b, See Attachment 3*
Ib A-2-a Ic See Attachment 3*
2 See Attachment 3*
3 See Attachment 3*
4 See Attachment 3*
5 A-3-a and A-3-b 6
A-1-b, See Attachment 3*
7 A-6-a, See Attachment 3*
8 A-6-b 9
See Attachment 3*
10 A-7-b 11 A-7-c 12 A-2-c 13 A-4-a 14 A-7-a, See Attachment 3*
15 A-1-a 16 A-9-a, See Attachment 3*
17a A-5-c 17b A-5-a 17c A-5-d 17d A-5-b 18a A-8-b 18b A-8-b 18c A-8-a 18d A-8 <i 18e A-8-c 19 A-10-a, See Attachment 3' 20 See Attachment 3* - VYNPS EAL NUREG-0654 Appendix 1 Deviations & Clarifications 1
1
BVY 99-94/ Attachment 2/Page 3 Vermont Yankee Nuclear Power Station NUREG-0654 to AP 3125 Cross-Reference NUREG-0654 IC AP 3125 EAL Site Area Emergency 1
S-2-b 2
S-2-a, S-2-b 3
See Attachment 3'*
4 S-3-a 5
See Attachment 3*
6 S-6-a, See Attachment 3*
7 S-6-b 8
S-7-b 9
S-7-c 10 S-2-c, S-2-d I1 S-4-a 12 S-7-a, See Attachment 3*
i 13a S-1-a, See Attachment 3*
)
13b S-1-a, See Attachment 3*
13c See Attachment 3*
14 S-9-a 15a S-5-c 15b S-5-a 15c S-5-b 16a S-8-b 16b S-8-a, S-8-b 16c S-8-c 17 S-10-a 18 S-8-d, See Attachment 3' - VYNPS EAL NUREG-0654 Appendix 1 Deviations & Clarifications
1 BVY 99-94/ Attachment 2/Page 4
)
Vermont Yankee Nuclear Power Station NUREG-0654 to AP 3125 Cross-Reference i
NUREG-0654 IC AP 3125 EAL General Emergency la G-1-a, See Attachment 3*
j lb G 1-a, See Attachment 3*
2 G-2-a, See Attachment 3*
3 G-9-a 4a G-10-a j
4b G-10-a de G-10-a 4d G-10-a 5a See Attachment 3*
Sb See Attachment 3*
Sc See Attachment 3*
Sd See Attachment 3*
Se See Attachment 3*
6a G-7-b 6b G-2-a, See Attachment 3*
6c G-2-a, See Attachment 3*
6d G-7-a 7
G-10-a l
- - VYNPS EAL NUREG-0654 Appendix 1 Deviations & Clarifications l
l
Docket No. 50-271 BVY 99-94 Vermont Yankee Nuclear Power Station NUREG-0654 Appendix 1 Deviations &
Clarifications i
BVY 99-94/ Attachment 3/Page 1
'Ihis awh= ant identifies and clarifies any deviations taken in AP 3125 from the generic NUREG-0654 Appendix 1 Initiating Conditions (ICs). Each NUREG-0654 IC number is listed in the same order as they are presented in Appendix 1 of NUREG-0654. If no deviation is taken to an IC, "None" is listed next to its corresponding IC number.
UNUSUAL EVENTS 1
Per USNRC EPPOS No.1, the NRC concludes that this EAL can be removed from the classification scheme provided that the scheme includes equivalent ICs for Unusual Event
- 5, Alert #5 and SAE #1. 'Ihe VY EALs have equivalent EALs for each of the specified NUREG-0654 ICs. 'Therefore this EAL can be deleted.
2 None 3a
'Ihe AEOG Hi Rad alann setpoint is based on the Technical Specification limit of 0.16 Ci/sec at the 30 minute delay point. Per Technical Specifications, this setpoint "prosides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Pan 100 in the event this effluent is inadvertently discharged directly to the environment without treatment." Note that the Technical Specification limit will result in declaration of an Unusual Event significantly before the NUREG example value of 500,000 pci/sec, and is therefore more restrictive.
3b Nec 3c
" Failed fuel monitor (PWR) indicates increase greater than.1% equivalent fuel failures" is not applicable since VY is a BWR.
4 VY design (i.e., Boiling Water Reactor) precludes the need for fuel or coolant temperature l
Technical Specification limits. Limiting Safety System Settings preclude these condition L
from occurring unless a failure to scram condition results. This would result in declaration of at least an Alert per AP 3125 EAL A-7-c.
USNRC EPPOS No. I states that licensees who eliminate this IC should incorporate an IC for " inability tc reach required shutdown within Technical Specifications limits." Hence EAL IC U-7-a USNRC EPPOS No. I funher states that licensees must maintain or add l
Unusual Event ICs for coolant activity exceeding Technical Specifications (EAL No. U a) and for RCS leakage exceeding Technical Specifications (EAL No. U-3-a).
Per the above referenced document, no emergency declaration is necessary if an LCO is satisfied within the time limit of the Technical Specification Action Statement. Operation within the boundaries of the Technical Specifications, including the specified action statements and restoration times, represents an analyzed and approved situation.
Therefore, it is concluded that an emergency condition exists only if operation occurs outside these boundaries; that is, if the required mode change or plant shutdown is not completed in the specified Technical Specifications Action Statement.
5 None l
IWY 99-94/ Attachment 3/Page 2 6
As stated in NUMARC, " Questions and Answers - Methodology for Development of Emergency Action Levels, NUMARC/NESP-007 Rev. 2", dated June 1993 in response to the question ".. should an emergency declaration be made if an SRV is stuck open or cycling?...," the above reference 1) stated that the answer depends upon other events. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration is not appropriate. However, if the fuel is damaged and the relief valve is allowing the fission products to escape into containment, a loss of the reactor coolant system (RCS) boundary has occurred.
For a stuck open or cycling SRV, Technical Specifications for suppression pool temperature and/or level may direct operators to shut down the plant. This is true for Vermont Yankee (see Technical Specification section 4.7). The above reference goes on to state that this condition is within the analyzed operating envelope of the plant and does not represent a degradation in the level of safety. This is also true for Vermont Yankee (see FSAR section 14.5.4.2, " Inadvertent Opening of a Relief Valve or Safety Valve").
Specifically, VY's FSAR states that no fuel damage results from this transient. Because pressure decreases throughout the transient, the nuclear system process barrier is not threatened by high internal pressure. The small amounts of radioactivity discharged with the steam are contained with the primary containment. The situation is not significantly different, from a radiological viewpoint, than that normally encountered in cooling the reactor using the relief valves to remove decay heat.
Failure of the SRV to close with reactor coolant temperature ;t 212 *F would be cater,orized as a LOSS of the REACTOR COOLANT SYSTEM BARRIER per AP-I 3125, " Fission Product Barrier Matrix." Coupled with other indications of multiple failed fuel barriers, a GENERAL EMERGENCY would be declared per EAL No. G-2-a, " Loss of 2 of 3 fission product barriers with the potential loss of the tidrd."
7 None 8
USNRC EPPOS No. I states that licensees who climinate this IC should incorporate an IC for " inability to reach required shutdown within Technical Specifications limits." Hence EAL IC U-7-a. USNRC EPPOS No. I further states that licensees must maintain or add Unusual Event ICs for coolant activity exceeding Technical Specifications (EAL No. U a) and for RCS leakage exceedmg Technical Specifications (EAL No. U-3-a).
Per the above referenced document, no emergency declaration is necessary if an LCO is satisfied within the time limit of the Technical Specification Action Statement. Operation within the boundaries of the Technical Specifications, including the specified action statements and restoration times, represents an analyzed and approved situation.
Therefoie, it is concluded that an emergency condition exists only if operation occurs outside these boundaries; that is, if the required mode change or plant shutdown is not completed in the specified Technical Specifications Action Statement.
9 USNRC EPPOS No. I states that licensees who eliminate this IC should incorporate an IC for " inability to reach required shutdown within Technical Specifications limits." Hence EAL No. U-7-a. USNRC EPPOS No. I further states that licensees must maintain or add I
1
IWY 99-04/ Attachment 3/Page 3 Unusual Event ICs for coolant activity exceeding Technical Specifications (EAL No. U a) and for RCS leakage exceeding Technical Specifications (EAL No. U-3-a).
Per the above referenced document, no emergency declaration is necessary if an LCO is satisfied within the time limit of the Technical Specification Action Statement. Operation within the boundaries of the Technical Specifications, including the specified action statements and restoration times, represents an analyzed and approved situation.
Therefore, it is concluded that an emergency condition exists only if operation occurs outside these boundaries; that is, if the required mode change or plant shutdown is not completed in the specified Technical Specifications Action Statement.
10 None 11 USNRC EPPOS No. I states that licensees may remove the significant loss of assessment
"., all meteorological instrumentation from this part of NUREG-0654 IC#11 due to the.
shift in emphasis from the classification based on dose assessment to classification based on plant conditions." llowever, licensees must maintain EALs for loss communications (EAL No. U-7-c) and loss ofindication or alarms (EAL No. U-7-b, A-7-a, and S-7-a).
For loss ofindications or alarms, USNRC EPPOS No. I also states that licensees may use the technical bases for NUMARC/NESP-007 ICs SU3, SA4 and SS6 to enhance their classification schemes provided that all of these ICs are implemented. EAL No. U-7-b reflects NUMARC IC SU3 bases.
12
" Attempted entry" and " attempted sabotage" are examples of security threats, and are not specifically contained in the initiating event. The condition of"affecting safe plant operation" is contained to conform to the definition of Unusual Event.
13a None 13b Nonc 13c None 13d None 14a The condition of" unusual aircraft activity over facility" has not been incorporated into EAL No. U-8-b. Unusual aircraft activity is difficult to define and is better addressed under EAL No. U-10-a "Other Conditions."
14b Train derailment near or on-site in and ofitself does not represent sufficient cause for declaration of an Unusual Event. This condition is adequately addressed under EAL U 4-b.
14c None 14d None
BVY 99-94/ Attachment 3/Page 4 14e.
Rapid shutdowns due to turbine rotating equipment failures is not a cause for declaration because a safe plant shutdown is achieved. The EAL has been changed to the more specific conditions cited in NESP-007 HUI based on NRC regulatory analysis for RG 1.101, Unusual Event Item 14.
15 USNRC EPPOS No. I states that licensees who eliminate this IC should incorporate an IC for " inability to reach required shutdown within Technical Specifications limits." Hence EAL No. U-7-a. USNRC EPPOS No. I further states tha. * ::nsees must maintain or add Unusual Event ICs for coolant activity exceeding Technical Specifications (EAL No. U a) and for RCS leakage exceeding Technical Specifications (EAL No. U-3-a).
Per the above referenced document, no emergency declaration is necessary if an LCO is satisfied within the time limit of the Technical Specification Action Statement. Operation within the boundaries of the Technical Specifications, including the specified action 4
statements and restoration times, represents an analyzed and approved situation.
Therefore, it is concluded that an emergency condition exists only if operation occurs outside these boundaries; that is, if the required mode change or plant shutdown is not completed in the specified Technical Specifications Action Statement.
EAL No. U-10-a provides the Emergency Director the latitude to declare based on other 3
conditions not specifically addressed in Categories 1 through 9.
l 16 Per USNRC EPPOS No.1, Emergency Preparedness Position (EPPOS) on Acceptable Deviations from Appendix 1 of NUREG-0654 Based Upon the Staffs Regulatory Analysis of NUMARC/NESP-007, " Methodology for Development of Emergency Action Levels",
dated 6/1/95, this event does not meet the threshold of the emergency class and is not a precursor to a more serious event. The event is reportable in accordance with 10CFR50.72 as a non-emergency.
17 This IC is not applicable since VY is a BWR.
ALERTS la Air ejector offgas radiation levels greater than the nominal hi-hi setpoint is indicative of significant fuel cladding failure and is consistent with the Alert EAL of 300 pCi/gm I-131 equivalent coolant activity. The AEOG Hi-Hi radiation alarm (set at a value corresponding to 0.15 Ci/sec)is well below the NUREG-0654 value of 5 Ci/sec, and complies with the requirements of Technical Specifications.
Ib None Ic Applicable only to PWRs.
2 Applicable only to PWRs.
3 Applicable only to PWRs.
4 The only MSIV malftmetion which could possibly cause or be associated with leakage from a steam line break would be a failure to close. Since this event is more appropriately
I BVY 99-94/ Attachment 3/Page 5 included in EAL No. S-3-a, a Site Area Emergency would be declared. There are no steam line break conditions which could occur which involve an MSIV malfunction and result in leakage unless the valve failed to shut (which is a Site Area Emergency). Any other steam line break which results in leakage would not be associated with an MSIV malfunction.
5 None 6
"Which could require offsite impact assessment" has been added to make the condition more consistent with the definition of an Alert, and to more clearly indicate when declaration is desired. The wording "off-site impact" relates to the release potential defined in the ALERT class description contained in NUREG 0654. This is further defined as any releases expected to be limited to small fractions of the EPA Protective Guideline exposure levels, as set forth in EPA-400.
7 Fifteen minutes was included as a threshold to exclude emergency declaration due to transient or momentary power losses. Mode applicability was limited to cold shutdown, refueling, or defueled mode because the significantly reduced decay heat, lower coolant temperature and RPV pressure provides additional time to restore one of the emergency busses relative to that available when the plant is hot or at power.
8
. None 9
Applicable only to PWRs.
10 None 11 None 12 None 13 None 14 USNRC EPPOS No. I states that licensees may use the technical bases for NUMARC/NESP-007 ICs SU3, SA4 and SS6 to enhance their classification schemes provided that all of these ICs are implemented. EAL No. A-7-a reflects NUMARC IC SA4 bases.
15 Nonc 16
" Threat" rather than npromise" is specified since it is more generic to VY "..the potential of affecting safe plant operation" specified to more clearly relate to the Alert definition.
I 17a None 17b None 17c None
BVY 99-94/ Attachment 3/Page 6 17d None 18a None 18b Nonc 18c-None 18d None 18e None 19 Rather than placing key emergency persosel on standby, VY procedures require that these personnel activate the TSC and EOF upa declaration of an Alen. Therefore, there is no need for a " standby" statement in the initiating event.
20 Per OP 3126 and EAL No. S-8-d, declaration of a Site Area Emergency is declared immediately when the decision is made to evacuate the control room. Since this is a more conservative approach, the intent of the NUREG example is met. The 15 minute criteria has not been used because of the difficulty involved in monitoring time while establishing control of shutdown systems and the difficulty of exactly defming when " control" has been established.
Site Area Emergencies 1
None 2
None 3
Applicable only to PWRs.
4 None 5
Applicable only to PWRs.
6 Mode applicability was limited to hot conditions to be consistent with EAL No. A-6-a.
See Alen 7.
7 None 8
None 9
None 10 None 1I None
BVY 99-94/ Attachment 3/Page 7 12 USNRC EPPOS No. I states that licensees may use the technical bases for NUMARC/NESP-007 ICs SU3, SA4 and SS6 to enhance their classification schemes provided that all of these ICs are implemented. EAL No. S-7-a reflects NUMARC IC SS6 bases.
13a As noted in INS 9507_02, "Re-evaluate Emergency Action Levels (EALs) for Possible Revision as Detailed on pages 3 &4 ofInspection Report", the integrated dose thresholds proposed in EAL No. S-1-a versus the dose rate thresholds in NUREG-0654, Appendix 1, are more consistent with the integrated doses provided in EPA-400. It states further that whole body and child thyroid doses are no longer utilized to evaluate protective actions per EPA-400. The total effective dose equivalent (TEDE) and committed dose equivalent (CDE) to the thyroid should be applied.
13b See 13b above.
13c Exceedmg EPA PAG dose thresholds appropriately requires declaration of a General Emergency per EAL No. G-1-a.
i 14 None 15a None 15b None 15c None 16a None 16b None 16c Nonc 17 None 18 The 15 minute criteria has not been used because of the difficulty involved in monitoring time while establishing control of shutdown systems and the difficulty of exactly defining when " control" has been established. The Site Area Emergency is declared per OP 3126 whenever the decision is made to evacuate the control room.
General Emergencies la As noted in INS 9507_02, "Re-evaluate Emergency Action Levels (EALs) for Possible Revision as detailed on pages 3 &4 ofInspection Report" the integrated dose thresholds proposed in EAL G-1-a, versus the dose rate thresholds in NUREG-0654, Appendix 1, are more consistent with the integrated doses provided in EPA-400. It states further that whole body and child thyroid doses are no longer utilized to evaluate protective actions per EPA-400. The total effective dose equivalent (TEDE) and committed dose equivalent (CDE) to the thyroid should be appied. According to EPA-400, the specified dose levels require evacuation of personnel, and thus, a General Emergency classification.
BVY 99-94/ Attachment 3/Page 8 Specifying " actual meteorological conditions" is felt to be unnecessary since a projected dose of the same magnitude wiy also result in classification at this EAL.
Ib See la above 2
Per NRC inspection report 50-271/92-14, Vermont Yankee was requested to evaluate this General Emergency (GE) Initiating Condition (IC) for additional indicators of fission product barrier integrity. Examples provided by the NRC were " containment pressure approaching design", and " oxygen / hydrogen mixture approaching combustible / explosive mixtures."
Fission Product Barrier Matrix to AP-3125 was created, which represents a matrix for evaluating the status of the three fission product barriers (i.e., fuel clad, reactor coolant system, & primary containment). This matrix is modeled after a similar concept proposed in NUMARC's " Methodology for Development of Emergency Action Levels",
NUMARC/NESP-007, Rev. 2, dated January 1992.
3 None 4a None 4b None 4c None 4d None 5a Applicable only to PWRs.
Sb Applicable only to PWRs.
Sc Applicable only to PWRs.
5d Applicable only to PWRs.
Se Applicable only to PWRs.
6a None 6b -
The conditions defmed by this IC are adequately addressed under General Emergency IC 2 (EAL No. G-2-a) based on fission product barrier loss.
6c The conditions defmed by this IC are adequately addressed under Gr neral Emergency IC 2 (EAL No. G-2-a) based on fission product barrier loss.
6d None 7
None