ML031290239
| ML031290239 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/30/2003 |
| From: | Public Service Enterprise Group |
| To: | David Pinckney Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML031290239 (14) | |
Text
Document Transmittal Form TO : NRC C/O PINCKNEY, DAVID ID: HECGO065 DOCUMENT CONTROL DESK WASHINGTON, DC 20555 Date: 4/30/03 Please update your controlled set of documents with the following documents:
Document ID Revision Status Quantity Format RecNo PRCIHECG-HECG-TOC-BASISO000 19 A 1 H 165889 PRCIHECG-SECT.08.1 (BASIS)1000 2 A 1 H 165932 ALL PAGES of this acknowledgement receipt must be returned to Document Management PSEG Nuclear PO Box 236 Hancocks Bridge, NJ 08038 MC N04 Your signature below verifies that:
(1) the above documents have been filed and superseded documents have been removed and destroyed or clearly marked as obsolete.
(2) the mailing address and copyholder information are correct or corrections have been identified on this transmittal.
Dj Place checkmark here to be removed from controlled distribution Signature: Date:
- ~C qD5
HOPE CREEK GENERATING STATION EVENT CLASSIFICATION GUIDE TECHNICAL BASIS April 30, 2003 CHANGE PAGES FOR REVISION #19 The Table of Contents forms a general guide to the current revision of each section and attachment of the Hope Creek ECG Technical Basis. The changes that are made in this TOC Revision
- 19 are shown below.
- 1. Check that your revision packet is complete.
- 2. Add the revised documents.
- 3. Remove and recycle the outdated material listed below.
ADD REMOVE Pages Description Rev. Pages Description Rev.
ALL TOC 19 All TOC 18 All Section 8.1 02 All Section 8.1 01 HC-ECG TB 1 of 1
PSEG Internal Use Only p Technical Basis NEELO~ T.O.C.,
HOPE CREEK ECG TECHNICAL BASIS CONTROL # HEC6C5 P
TABLE OF CONTENTS/SIGNATURE PAGFCopy SECTION TITLE REV # PAGES DATE T.O.C. Table of Contents/Signature Page 19 4 04/30/03 Introduction and Usage 00 3 01/21/97 ii Glossary of Acronyms & Abbreviations 00 5 01/21/97 1.0 Fuel Clad Challenge 01 9 06/14/01 2.0 RCS Challenge 00 8 01/21/97 3.0 Fission Product Barriers (Table) 3.1 Fuel Clad Barrier 03 13 02/01/02 3.2 RCS Barrier 02 18 02/01/02 3.3 Containment Barrier 04 16 11/11/02 4.0 EC Discretion 00 8 01/21/97 5.0 Failure to SCRAM 00 10 01/21/97 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release 00 44 01/21/97 6.2 Liquid Effluent Release 00 4 01/21/97 6.3 In - Plant Radiation Occurrences 00 6 01/21/97 6.4 Irradiated Fuel Event 01 8 11/15/01 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 01 18 05/12/P7 7.2 Loss of DC Power Capabilities 00 5 01/21/97 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 02 8 04/30/03 8.2 Loss of Overhead Annunciators 00 8 01/21/97 8.3 Loss of Communications Capability 00 4 01 :1/97 8.4 Control Room Evacuation 00 4 01/21'/7 8.5 Technical Specifications 00 2 01/21/97 9.0 Hazards - Internal/External 9.1 Security Threats 02 9 02/01/02 9.2 Fire 01 6 02/01/01 9.3 Explosion 01 5 02/01/01 9.4 Toxic/Flammable (Gases 02 13 11/11/02 9.5 Seismic Event 02 4 11/11/02 9.6 High Winds 01 7 02/01/01 9.7 Flooding 01 5 02/01/01 9.8 Turbine Failure/V ,hicle Crash/ 01 7 02/'1 /01 Missile Impact 9.9 River Level 00 4 01/91/97 HCGS Rev. 19
PSEG Internal Use Only HC EAL Technical Basis T.O.C.
Pg. 2 of 4 HOPE CREEK ECG TECHNICAL BASIS TABLE OF CONTENTS/SIGNATURE PAGE SECTION TITLE REV # -
PAGES DATE 10.0 Reserved for future use 11.0 Reportable Action Levels (RALs) 11.1 Technical Specifications 02 7 01/23/01 11.2 Degraded or Unanalyzed Condition 03 4 02/28/02 11.3 System Actuation 05 7 04/19/02 11.4 Personnel Safety/Overexposure 01 8 01/23/01 11.5 Environmental/State Notifications 01 4 01/23/01 11.6 After-the-Fact 02 1 02/28/02 11.7 Security/Emergency Response 04 5 02/28/02 Capabilities 11.8 Public Interest 01 01/23/01 11.9 Accidental Criticality/ 02 8 01/23/01 Special Nuclear Material /
Rad Material Shipments - Releases 11.10 Voluntary Notifications 01 2 01/23/01 HCGS Rev. 19
PSEG Internal Use Only HC EAL Technical Basis T.O.C.
Pg. 3 of 4 REVISION
SUMMARY
Biennial Review Performed: Yes _ No X 8.1 .3.b has a clarification made for the condition of "If Primary Containment Instrument Gas/Instrument Air (PCIGIIA) is lost due to LOCA conditions and the MSIVs are drifting shut, consider the Main Condenser capability lost (imminent <2 hours) for the purpose of this EAL." This follows the direction given in the ECG usage section.
HCGS Rev. 19
PSEG Internal Use Only HC EAL Technical Basis T.O.C.
Pg. 4 of 4 SIGNATURE PAGE Prepared By: F. J. Hughes 04/10/03 Date Section/Attachments Revised d1AS (List Non-Editorial Only - Section/Attachments) Date Reviewed By: N/A_
I OCFR50.54q Effectiveness Reviewer Date Reviewed By: N/A Department Manager Date Reviewed By: N/A Manager - Licensing Date (Reportable Action Level (S tio n d associated Attachments marked by "L")
Reviev ved By: (. JJz.
I Emergd icy Preparedness Manager 11i
' D~te Reviewed By: N/A Manager - Quality Assessment - NBU Date (If Applicable)
SORC Review and Station Approvals N/A N/A_
Mtg. No. Hope Creek Chairman Vice President - Nuclear Operations Date Date Effective Date of this Revision: - 30 ^Date Date HCGS Rev. 19
HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability PSE&G ALERT - 8.1.2 CONTROL IC Inability to Maint ain the Plant in Cold Shutdown COPY # H C&( -5 EAL Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions AND EITHER one of the following occur:
- RCS Temperature has risen to > 200 TF (Excluding a < 15 minute rise > 200 TF with a heat removal function restored)
- An UNCONTROLLED temperature rise is RAPIDLY approaching 200 TF (with NO heat removal function restored)
OPERATIONAL CONDITION - 4,5 BASIS Loss of Decay Heat Removal capabilities necessary to maintain Cold Shutdown conditions could potentially lead to core damage is corrective actions are not implemented. Declaration of an Alert is warranted when ALL Technical Specification required systems are not available to provide Decay Heat Removal functions and cannot be restored to prevent boiling in the core.
The specification of an RCS temperature rise, rather than specific equipment failures, recognizes the potential for long heatup times providing adequate time for restoration of some form of alternate cooling.
The statement "Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions" is intended to represent a complete loss of functions available, or an inadequate ability, to provide core cooling during the Cold Shutdown and Refueling Moues, including alternate decay heat removal methods. This EAL allows for actions taken IAW OP-AB.ZZ-0142, Loss of Shutdown Cooling (Abnormal Operating Procedure) to reestablish RHR in the Shutdown Cooling Mode or provide for an EAL - 8.1.2 Rev. 02 Page 1 of 3
HCGS EAL/RALTechnical Basis alternate methods of decay heat removal, with the intent of maintaining RCS temperature below 200 0 F.
For loss of an in-service Decay Heat Removal system with other decay heat removal methods available, actions taken to provide for restoration of a decay heat removal function may require time to implement. If the event results in RCS temperature "momentarily" (for less than 15 minutes) rising above 200'F with heat removal capability restored, Emergency Coordinator judgment will be required to determine whether heat removal systems are adequate to prevent boiling in the core and restoration of RCS temperature control. Momentary (not to exceed 15 minutes) unplanned excursions above 2000F, when alternate decay heat removal capabilities exist, should not be classified under this EAL.
NRC analysis has shown that specific sequences can result in core uncovery within 15 to 20 minutes and severe core damage within an hour after decay heat removal capability has been lost.
Unplanned is defined as a condition that is not due to scheduled operations or maintenance activities, in which an RHR system is intentionally removed from service.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on inability to maintain RPV Water level above the Top of the Active Fuel, or rising Radiological Releases.
DISCUSSION The Residual Heat Removal (RHR) system provides the normal method for decay heat removal operating in the Shutdown Cooling Mode. With RHR unavailable for shutdown cooling operation, (including the loss of SACS and/or service water which supply cooling water to the RHR heat exchangers), alternate decay heat removal system can be aligned to control decay heat.
An unavailability of these systems, can result in a gradual rise in RCS temperature to the values specified in this EAL. The rate of rise in coolant temperature would be dependent on the amount of decay heat present. The threshold for this EAL is the RCS temperature transition value between Operational Conditions 4 and 3.
Procedural guidance is provided to establish an alternate method of decay heat removal. These alternate methods include: aligning Reactor Water Cleanup system (RWCU), with maximum RACS aligned to the Non-Regenerative Heat Exchanger; aligning Condensate Transfer via the ECCS injection lines; aligning RPV Head Spray with RPV Water Level established above
+80"; maximizing Fuel Pool Cooling if the RPV head is removed and the reactor cavity flooded; using the "C" RHR pump crosstied to the "A" RHR loop.
EAL- 8.1.2 Rev. 02 Page 2 of 3
HCGS EAL/RALTechnical Basis If these alternate means are unavailable, or ineffective, decay heat removal must be accomplished by feed-and-bleed using ECCS systems and discharging steam to the Suppression Pool via the SRVs.
DEVIATION None REFERENCES NUMARC NESP-007, SA3 NUMARC Questions and Answers, June 1993, "System Malfunction Question #6b" HC.OP-AB.ZZ-0142 (Q), Loss of Shutdown Cooling HC.OP-EO.ZZ-0I01 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-O102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NEDO-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications Sections 3/4.3,3/4.4.9,3/4.7.1,3/4.7.2 EAL- 8.1.2 Rev. 02 Page 3 of 3
HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY - 8.1.3.a IC Loss of Reactor Water Level that has or will Uncover Fuel in the Reactor Vessel EAL OPERATIONAL CONDITION - 4,5 BASIS Reactor Water Level reaching -161" (Top of Active Fuel) indicates a loss of core submergence.
Without core submergence, the integrity of the fuel clad barrier can no longer be assured, even with the reduced decay heat levels in Cold Shutdown and Refuel. This event is classified based on reaching the Reactor Water level threshold (instead of being able to restore and maintain above the threshold) due to the potentially severe consequences of a loss of core submergence.
Since the design of the normal and emergency makeup systems should preclude this condition, an extreme challenge to their ability to provide core cooling by submergence has occurred.
Additionally, ECCS availability and Containment Integrity requirements may be relaxed under these Operational Conditions, thus classification at the Site Area Emergency level is warranted.
Barrier Analysis Fuel Clad Barrier has been potentially lost RCS Barrier has been lost.
ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on abnormal Radiological Releases.
EAL - 8.1.3.a Rev. 02 Page 1 of 2
HCGS EAL/RALTechnical Basis DISCUSSION Core Submergence ensures adequate core cooling. When RPV water level decreases to below Top of Active Fuel (TAF) the ability to effectively remove decay heat can no longer be guaranteed and the Fuel Cladding Barrier can no longer be considered intact. Sustained partial or total core uncovery can result in clad damage and a significant release of fission products to the reactor coolant. Sustained core uncovery can also result in a breach of the reactor vessel, or an unisolated intersystem LOCA with the RHR System.
DEVIATION None REFERENCES NUMARC NESP-007, SS5 HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control EAL- 8.1.3.a Rev. 02 Page 2 of 2
HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY- 8.1.3.b IC Complete Loss of Functions Needed to Achieve Cold Shutdown Conditions EAL Loss of Main Condenser capabilities, as evidenced by an inability to remove Decay Heat from the Reactor AND Loss of Torus capabilities as evidenced by EITHER one of the following:
- Entry into an Unsafe region of ANY of the following curves:
- Heat Capacity Temperature Limit (HCTL) Curve
- Heat Capacity Level Limit (HCLL) Curve
- Pressure Suppression Pressure (PSP) Curve
- SRV Tailpipe Level Limit Curve
- Insufficient SRV capacity to reduce RPV pressure OPERATIONAL CONDITION - 1, 2,3 BASIS A Complete Loss of decay heat removal systems required to ACHIEVE Cold Shutdown conditions from a Hot Shutdown condition, represents a significant challenge to the plant due to the failure of multiple systems designed for the protection of the public. Hence, declaration of a Site Area Emergency is warranted.
This EAL specifically includes a degradation of those plant systems required to ACHIEVE a Cold Shutdown condition. It does NOT include an inability to MAINTAIN a Cold Shutdown condition. The inability to MAINTAIN Cold Shutdown Conditions is specifically addressed by EAL 8.1.2. Hence, a Loss of RHR Shutdown Cooling is not included in this EAL.
This EAL includes a loss of Service Water or SACS capabilities, based on the effect a loss of these systems has on the ability to maintain Torus capabilities with the Safe Region of the EAL- 8.1.3.b Rev. 02 Page 1 of 3
HCGS EAL/RALTechnical Basis referenced EOP curves. Loss is defined as the systems being unavailable to perform their intended design function.
If Primary Containment Instrument Gas/Instrument Air (PCIG/IA) is lost due to LOCA conditions and the MSIVs are drifting shut, consider the Main Condenser capability lost (imminent <2 hours) for the purpose of this EAL.
In the case where the Main Condenser became isolated from the Reactor due to an MSIV Isolation, but the MSIV could be reopened by procedure, or Main Steam Line drains can control pressure, then a Loss of the Main Condenser capabilities has not occurred.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on loss of Fission Product Barriers or Radiological Releases.
DISCUSSION In this event, a loss of both the normal heat sink for the Reactor and an impending severe degradation of alternate heat removal capability to the Torus has occurred. Loss of the heat sink for the reactor when in a Hot Shutdown condition will limit the ability to maintain that Operational Condition, or to cooldown the reactor if required.
The Main Condenser can be lost for a variety or reasons; loss of Circulating Water, loss of the Turbine Control and/or Bypass Valve functions, Main Steam Line isolation, etc. With the Main Condenser not available and without the RHR System lined up in Shutdown Cooling Mode, decay heat must be removed from the RCS by HPCI, RCIC or the SRVs and be absorbed in the Suppression Pool (SP). Loss of the pressure control ability of the SRVs as indicated by the inability to reduce RPV pressure represents a loss of control of a major RCS parameter which could result in RPV overpressure conditions, or the inability to cooldown if Cold Shutdown is required.
The HCTL curve is defined as the highest Torus temperature at which initiation of RPV depressurization will not result in exceeding either the SP design temperature or the Primary Containment pressure limit before the rate of energy transfer from the RPV to the Primary Containment is beyond the capacity of the Containment Vent.
The HCLL curve is defined as the higher of either the elevation of the Containment downcomer opening or the lowest Torus level at which initiation of RPV depressurization will not result in exceeding the HCTL.
EAL - 8.1.3.b Rev. 02 Page 2 of 3
HCGS EAL/RALTechnical Basis Violation of either curve would require an immediate emergency depressurization, thus ensuring that the immediately present thermal energy in the RCS has been transferred to the Primary Containment while maintaining the Containment within design limits. This represents a serious potential threat to the Primary Containment Barrier.
DEVIATION The NUMARC IC associated with EAL SS4 suggests that the IC should include a Complete Loss of Functions needed to achieve or maintain Hot Shutdown. The NUMARC basis includes both reactivity control and decay heat removal. At Hope Creek, as with all other BWRs, the operator action of placing the Reactor Mode Switch in the Shutdown position that results in Control Rod inserting into the core such that the Reactor will remain shutdown under all conditions without boron, places the Reactor in a Hot Shutdown condition. No additional actions are required to maintain the Reactor in this condition.
Systems are required and additional operator actions are required to achieve Cold Shutdown conditions. Based on this, Hope Creek has modified the NUMARC IC for SS4 to apply specifically to a total loss of decay heat removal, since reactivity control concerns are addressed under the ATWS Section. This IC and EAL are consistent with the requirements for declaration of a Site Area Emergency.
REFERENCES NUMARC NESP-007, SS4 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-I0O1 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NEDO-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 3/4.1.3, 3/4.1.5 EAL- 8.1.3.b Rev. 02 Page 3 of 3