ML20065N339
| ML20065N339 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 09/23/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20063N686 | List: |
| References | |
| 22A7007, NUDOCS 8210220174 | |
| Download: ML20065N339 (259) | |
Text
{{#Wiki_filter:- _. GESSAR II 22A7007 238 NUCLEAR ISLAND Rzv. 7 INSTRUCTIONS FOR FILING AMENDMENT NO. 7 Remove and insert the pages listed below. Remove Insert Appendix 1A and 1B 1A-xi, IA-xii, IA-xiii/1A-xiv 1A-xi, IA-xii, lA-xiii, 1A-xiv, 1A-xv/1A-xvi, 1A.8-3, 1A.8-4, 1A-xv/1A-xvi, lA.8-3, 1A.8-4, 1A.9-1, 1A.19-2, lA.19-5, 1A.9-1, 1A.19-2, 1A.19-5, 1A.19-8, 1A.19-9, 1A.23-1, lA.19-8, 1A.19-9, 1A.23-1, 1A.23-2, IA.23-3, 1A.23-5/ 1A.23-2, 1A.23-3, 1A.23-5/ , 1A.23-6, 1A.24-2, 1A.24-4, 1A.23-6, 1A.24-2, 1A.24-4, 1A.39-2, 1A.58-3, 1A.61-1/ 1A.39-2, 1A.58-3, 1A.61-1/ 1A.61-2, 1A.67-3, 1A.68-2, 1A.61-2, 1A.67-3, 1A.68-2, 1A.71-1, 1A.74-1/1A.74-2, 1A.71-1, 1A.74-1/1A.74-2, 4 1A.80-2, 1A.80-4, lAA-5, 1A.80-2, 1A.80-4, 1A.80-5, 1AA-8, 1AA-9, 1AA-11, 1AA-15/ 1AA-5, 1AA-8, 1AA-9, lAA-11, lAA-16, lAA-18, IAA-20, 1AA-15/1AA-16, lAA-18, 1AA-20, 1AA-26, IAA-27, 1AA-28, 1AA-29, lAA-26, IAA-27, IAA-28, 1AA-29 1AA-30, 1AA-31/1AA-32, 1AA-33, 1AA-30, 1AA-31/1AA-32, 1AA-33, 1AA-34, 1AA-35, 1AA-36, lAA-37, 1AA-34, 1AA-35, IAA-36, lAA-37, 1AA-38, 1AA-39, lAA-40, IAA-41/ 1AA-38, lAA-39, 1AA-40, lAA-41/ lAA-42, 1AA-44, 1AB-3/1AB-4, 1AA-42, IAA-44, 1AB-3/1AB-4, . LAB-7, IAB-8, LAC-4, 1AC-5, LAB-7, 1AB-8, LAC-4, 1AC-5, 1AC-7/1AC-8, IAC-18, LAC-19, 1AC-7/1AC-8, 1AC-18, 1AC-19, IAC-21, IB-i, IB-1, 18-2, LAC-21, IB-i, IB-1, 18-2, 18-5, IB-8, 1B-12, 1B-17, 18-5, 1B-8, IB-12, 1B-17, 1B-19, 1B-20, 18-23, and B-28 IB-19, 1B-20, 1B-23, and 18-28 Appendix 38 38-47/3B-48 3B-47/3B-48 Chapter 4 4.1-i, 4.1-11, 4.1-iii/4.1-iv, 4.1-1/4.1-ti, 4.1-5 through through 4.1-23/4.1-24, 4.3-111/ 4.1-12, 4.3-iii/4.3-iv, 4.3-iv, 4.3-1, 4.3-3, 4.3-4, 4.3-1, 4.3-3, 4.3-4, 4.3-6, 4.3-6, 4.3-10, 4.4-11, 4.4-12, 4.3-9a, 4.3-10, 4.4-11, 4.4-12, 4.4-13/4.4-14, and 4.4-19/4.4-20 4.4-13/4.4-14, and 4.4-19/4.4-20
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1 ! Appendix 4A l ! 4A-1/4A-ii, 4A-v/4A-vi, 4A.2-1/ 4A-i/4A-ii, 4A-v/4A-vi, 4A.2-1/ l 4A.2-2, 4A.3-1/4A.3-2, 4A.5-27, 4A.2-2, 4A.3-1/4A.3-2, 4A.6-27, l 4A.5-28, and 4A.5-29/4A.5-30 4A.6-28, and 4A.6-29/4A.6-30 Chapter 6 6.3-29, 6.3-31, 6.3-34, 6.3-42 6.3-29, 6.3-31, 6.3-34, 6.3-42, and 6.3-48 and 6.3-48 B210220174 820923 l PDR ADOCK 05000447 A PDR ~1- ! 1m/E09275-1 Amendment 7 September 23, 1982
GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 7 Remove Insert 15.1-v/15.1-vi, 15.1-14, 15.1-v/15.1-vi, 15.1-14, 15.1-15, 15.1-17, 15.1-18, 15.1-15, 15.1-17, 15.1-18, 15.1-19, 15.1-20, 15.1-21 15.1-19, 15.1-20, 15.1-21, 15.1-22, 15.1-23, 15.1-24, 15.1-22, 15.1-23, 15.1-24, 15.1-25, 15.1-26, 15.2-xi, 15.1-25, 15.1-26, 15.2-ix, 15.2-xii, 15.4-v/15.4-vi, 15.2--x, 15.4-v/15.4-vi, 15.4-vii/15.4-viii, 15.4-5, 15.4 yii/15.4-viii, 15.4-5 15.4-9, 15.4-12, 15.4-16, 15.4-9, 15.4-12, 15.4-16, 15.4-35, 15.4-36, 15.4-37, 15.4-25, 15.4-26, 15.4-27 15.4-38, 15.4-39, 15.4-40, 15.4-28, 15.4-29, 15.4-30, 15.4-32, 15.4-43, 15.4-44, 15.4-32, 15.4-33, 15.4-34, 15.4-45, 15.4-46, 15.4-37, 15.4-35, 15.4-36, 15.4-37, 15.4-38, 15.4-39, 15.4-40, 15.4-38, 15.4-39, 15.4-40, 15.4-51, 15.4-52, 15.4-53/ 15.4-41, 15.4-42, 15.4-43/ 15.4-54, 15.4-55, 15.4-56, 15.4-44, 15.4-45, 15.4-46, 15.4-57, and 15.4-58 15.4-47, and 15.4-48 Appendix 15D 15D.2-2, 15D.2-3, 150.2-4, 15D.2-2, 15D.2-3, 15D.2-4, 150.2-5, 15D.2-7, 150.2-11, 150.2-5, 150.2-7, 15D.2-11, 150.2-13, 150.2-15, 15D.2-17, 15D. 2-13, 15D. 2-15, 15D . 2- 17, 15D.2-20, 15D.2-21, 150.2-27, 150.2-20, 15D.2-21, 150.2-27, 150.2-39, 15D.2-40, 150.2-45, 150.2-39, 150.2-40, 15D.2-45, 150.2-47, 15D.2-48, 15D.2-49, 150.2-47, 15D.2-48, 150.2-49, 150.2-50, 15D.2-51, 15D.2-52, 15D.2-50, 15D.2-51, 150.2-52, 15D.2-53, 150.2-71, 15D.2-74, 15D.2-52a, 150.2-53, 15D.2-71, 15D.2-78a/150.2-78b, 15D.2-81, 150.2-74, 150.2-78a/150.2-78b, 15D.4-8, 150.4-9, 15D.4-15/ 150.2-81, 150.4-8, 150.4-9, 15D.4-16, 15DA-14, 15DA-16, 15D.4-15/15D.4-16, 15DA-14, 15DA-34, 150A-39, 15DA-40, and 15DA-16, 15DA-34, 15DA-39, 15DA-41 15DA-40, and 15DA-41 0 Im/E09275-2 Amendment 7 September 23, 1982
_ _ _ _. _ - . . _ _ ._ - . . . _ _ _ . _ ~. .. GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 i. i CONTENTS (Continued) i SECTION TITLE PAGE 1A.79 CONTROL-ROOM HABITABILITY REQUIREMENTS 1A.79-1 t l (III.D.3.4) ! 1A.80 REFERENCES 1A.80-1
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\ j 1AA ATTACHMENT A TO APPENDIX 1A - COMPLIANCE 1AA.1-1 _ r
. OF 238 NUCLEAR ISLAND WITH NUREG-0737 - -
i II.B.2 , 1AA.1 NRC POSITION 1AA-1 , lAA.2
SUMMARY
OF SHIELDING DESIGN REVIEW 1AA-3 I 1AA.3 CONTAINMENT DESCRIPTION AND DEFINITION OF TERMS 1AA-7
- 1AA.3.1 Description of Primary / Secondary Containment 1AA.3.2 Vital Areas 1AA-9
- 1AA.3.3 Post Accident Operation 1AA-9 ,
l 1AA.4 DESIGN REVIEW BASES 1AA-11 _. l 1AA.4.1 Radioactive Source Term 1AA-11 I 1AA.4.2 Safety-Related Equipment Requiring j Qualification } 1AA.4.3 Qualification Duration of Safety-Related Equipment 1AA-11 1AA.4.4 Availability of Off-Site Power 1AA-12 , 1AA.4.5 Accidents Used as the Basis for the Specified Radioactivity Release 1AA-12 1AA.4.6 Environmental Qualification Conditions 1AA-14 l 1AA.5 RESULTS OF THE REVIEW 1AA-17 2
! 1AA.S.1 Systems Required Post-Accident 1AA-17 l l 1A-xi 1
16H12 f
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 CONTENTS (Continued) SECTION TITLE PAGE 1AA.5.1.1 Necessary Post-Accident Functions and Systems 1AA-17 1AA.5.1.2 Emergency Core Cooling Systems and Auxiliaries 1AA-18 1AA.S.l.3 Fission Product Removal and Control Systems and Auxiliaries 1AA-20 1AA.5.1.4 Combustible Gas Control Systems and Auxiliaries lAA-21 1AA.S.1.5 Instrumentation and Control Systems Auxiliaries 1AA-22 , 1AA.S.2 Environmental Conditions for Post-Accident Systems 1AA-22 1AA.S.2.1 Calculation Results 1AA-22 1AA.S.2.2 Discussion of Conservatism in Calculation 1AA-23 1AA.S.2.3 Discussion of Fuel Building Access Neds 1AA-25 1AA.6 CONCLUSIONS OF THE REVIEW 1AA-27 ] 1AB ATTACHMENT B TO APPENDIX 1A - TECHNICAL DESCRIPTION OF 238 NUCLEAR ISLAND POST-ACCIDENT SAMPLING STATION 1AB-1 1AB.1 PLANNED MODIFICATION 1AB-1 1AB.2 GAS SAMPLES 1AB-3 1AB.3 LIQUID SAMPLES 1AB-5 1AC ATTACHMENT C TO APPENDIX 1A - CONTAINMEh"I ISOLATION DEPENDABILITY 1AC.1 1AC.1 INTRODUCTION 1AC-1 1AC.2 238 NUCLEAR ISLAND ASSESSMENT 1AC-3 1AC.3 MODIFICATIONS TO MEET POSITION ITEM 3 1AC-5 1A-xii 16H13
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 I CONTENTS O (Continued) SECTION TITLE PAGE 1AC.3.1 RHR and Reactor Water Sample Lines LAC-6 1AC.3.2 RCIC Steam Supply Lines LAC-6 1AC.4 DISCUSSION OF CONTAINMENT PURGE DESIGN 1AC-9 1AC.5 CONCLUSION 1AC-15 1 1 !O l l 1A-xiii i 16H14 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 TABLES TABLE TITLE PAGE lAA-1 Required Environmental Conditions 1AA-29 , lAA-2 Radiation Source Comparison 1AA-30 _ lAA-3 Core Cooling Systems and Auxiliaries - Major Equipment lAA-31 1AA-4 Core Cooling Systems and Auxiliaries Valves and Instrument Transmitters lAA-33 1AA-5 Fission Product Removal and Control 1AA-35 1AA-6 Combustible Gas Control 1AA-38 1AA-7 Compartment Radiation Dose Rate vs. Time After Accident 1AA-8 Post Accident Radiation Exposure, RADS 1AA-40 1AA-9 Comparison of Calculated Exposures vs. Required Exposures for Post Accident Systems lAA-41 1AC.2-1 Essential / Nonessential Equipment LAC-17 l i I 1A-xlv h 16D1 l l
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 FIGURES FIodRE TITLE PAGE i 1A.24-1 Planned Modification for SRV Open/ Closed Monitor (Figure 5.1-3a) 1A.24-5 1A.59-1 A Typical RCIC Steam Line Break Detection Logic Diagram 1A.59-3 1A.59-2 Schematic Diagram of Time Delay Action to Preclude Spurious RCIC Isolation During System Start Sequence 1A.59-4 1A.67-2 Fuel Zone Common Reference Zaro 1A.67-4 lAA-1 Diesel Generator Building 1 Radiation vs. Time Following LOCA 1AA-43 1AA-2 ECCS Rooms - Dose Rate and Integrated Exposure, Post LOCA 1AA-44 1AB.1-1 Post-Accident Sampling Station LAB-11 () 1AB.3-1 General Arrangement Post-Accident Sample Station 1AB-13 1AB.3-2 Main Control Panel 1AB-14 ) 1AC-3-1 RHR and Reactor Water Sample Lines 1AC-23 1AC-3-2 RCIC Steam Supply Lines 1AC-24 i I r O i (_,/ 1A-xv/1A-xvi 16D2
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.8 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF FROCEDURES FOR TRANSIENTS AND ACCIDENTS (NUREG-0737 Item I.C.1) (Cont'd) Response (Cont'.d)
- 2. NEDO-24708A, Revision 1, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," December, 1980.
This report was issued via the letter from D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC) dated March 20, 1981.
- 3. BWR Emergency Procedure Guidelines (Revision 0) --
submitted in prepublication form June 30, 1980.
- 4. BWR Emergency Procedure Guidelines (Revision 1) -
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Issued via the letter from D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC) dated January 31, 1981.
- 5. BWR Emergency Procedure Guidelines (Revision 2) -
submitted in prepublication form June 1, 1982, Letter BWROG 8219 from T. J. Dente (BWR l Owners Group) to D. G. Eisenhut (NRC). i b. Adequacy of Submittals l l l The submittals described in Paragraph (a.) have I been discussed and reviewed extensively among the BWR Owners' Group, the General Electric Company, l and the NRC Staff. The NRC Staff has found I (NUREG-0737, p. I.C.1-3) that "the analysis and I j 1A.8-3 i \_- 125L1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.8 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS (NUREG-0737 Item I.C.1) (Cont' Response (Cont'd) guidelines submitted by the General Electric Company (GE) Owners' Group... comply with the requirements (of the NUREG-0737 clarifications)." In Reference 9, the Director of the Division of Licensing states, "we find the Emergency Procedure Guidelines acceptable for trial implementation (on six plants with applications for operating licenses pending)." GE believes that in view of these findings, no further detailed justification of the analyses or guidelines is necessary at this time. Emergency procedures developed from the emergency procedures guidelines will be prepared by each applicant and implemented prior to fuel loading. Section 15D.2.3 also addresses the emergency procedure guidelines with regard to their relation to severe accidents. The emergency procedures training program , will be made available for review by the NRC by the applicant. i O 1A.8-4 125L2
- GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 lA.9 SHIFT RELIEF AND TURNOVER PROCEDURES (NUREG-0737 Item I.C.2.) (Cont'd)
NRC Position
- The licensees shall review and revise as necessary the l
plant procedure for shift and relief turnover to assure the following:
- a. A checklist shhll be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete'and sign. The following items, as a minimum, shall be included in the checklist.
- 1. Assurance that critical plant parameters are within allowable limits (parameters and j() allowable limits shall be listed on the checklist).
- 2. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the .
control console. (What to check and criteria ! for acceptable status shall be included on l the checklist) . l l ! 3. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of time in i the degraded mode shall be compared with the i 1A.9-1 ! 125L3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.9 SHIFT RELIEF AND TURNOVER PROCEDURES (NUREG-0737 Item I.C.2) (Cont'd) NRC Position * (Cont'd) Technical Specifications action statement (this shall be recorded as a separate entry on the checklist).
- b. Checklist or logs shall be provided for completion by the offgoing and ongoing auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist); and O
- c. A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments) .
Response
The response to this requirement will be supplied by the applicant.
- This position statement is repeated from Reference 10 since it was not provided in detai in either NUREG-0660 or NUREG-0737 O
1A.v-2 125K10
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 r 1A.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 Item II.B.1) I . NRC Position Each applicant and licensee shall install reactor , coolant system (RCS) and reactor vessel head high point vents remotely operated from the control room. Although the purpose of the system is to vent noncondensible
- gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to
, 10 CFR Part 50, " General Design Criteria." The vent I system shall be designed with sufficient redundancy that assures a low probability of inadvertent or O irreversible actuation. Each license shall provide the following information concerning the design and operation of the high point l vent system: (1) Submit a description of the design, location, size, and power supply for the vent system along l with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe. The results of the analyses should demonstrate l compliance with the acceptance criteria of
- 10 CFR 50.46.
l l l 1A.19-1 132A1 i ._.
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 Item II.B.1) (Cont'd) NRC Position (Cont'd) (2) Submit procedures and supporting analysis for operator use of the vents that also include the information available to the operator for initiating or terminating vent usage.
Response
The capability to vent the 238 Nuclear Island reactor coolant system is provided by the safety relief valves and reactor coolant vent line as well as other systems. The capability of these systems and their satisfaction of Item II.B.1 is discussed below. The 238 Nuclear Island design is provided with nineteen power-operated safety-grade relief valves which can be manually operated from the control room to vent the reactor pressure vessel. The point of connection to the main steamlines which exit near the top of the vessel to these valves is such that accumulation of gases above that point in the vessel will not affect removal of gases from the reactor core region. These power-operated relief valves satisfy the intent of the NRC position. Information regarding the design, qualification, power source, etc., of these valves is provided in Subsection 5.2.2. 1A.19-2 132F1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 j 1A.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737-Item IT.B.1) (Cont'd) Response (Cont'd) i Under most circumstances, no selection of vent path is t 4 necessary-because the relief valves (as part of the ) automatic depressurization system), HPCS, and RCIC will
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function automatically in their designed modes to ensure adequate core cooling and provide continuous j venting to'the suppression pool. ; 1 Analyses of water inventory-threatening events with f i very severe degradations.of system performance have I been conducted. These were submitted by GE for the BWR Owners' Group to the NRC Bulletins and Orders Task . ! Force on November 30, 1979 (Reference 24). The funda-mental conclusion of those studies was that if only () one ECC system is injecting into the reactor, adequate core cooling would be provided and the production of
- 1. Je quantities of hydrogen was avoided. Therefore, ,
4 ! it is not desirable to interfere with ECCS functions-l to prevent venting. I The emergency procedure guidelines emphasize the use of l HPCS/RCIC as a first line of defense for inventory- ! threatening events which do not quickly depressurize the reactor. If these systems succeed in maintaining. 1 i inventory, it is desirable to leave them in operation ! until the decision to proceed to cold: shutdown is made. j Thus, the reactor will be vented via RCIC turbine steam 4 being discharged to the suppression pool. Termination , j of this mode of venting could also terminate inventory. This would necessitate makeup if the HPCS had failed also. >
- reactor depressurization via the SRV, which of' course-() is another means of venting.
i i 1A.19-5 i 132F2 i _ , - _ _ _ . - , - . _ . _ _ . . _ . _ _ . . . - ~ . , . - _ . , _ - - - . - _._ _ _ _ , _, .
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 Item II.B.1) (Cont'd) Response (Cont'd) If the HPCS/RCIC are unabla to maintain inventory, the emergency procedures guidelines call for use of ADS or manual SRV actuation to depressurize the reactor so that the low-pressure LPCI and/or LPCS systems can inject water. Thus, the reactor would be vented via the SRV to the suppression pool. Termination of this mode of venting is not recommended. It is preferrable to remain unpressurized; however, if inventory makeup requires HPCS or RCIC restart, that can be accomplished manually by the operator. It is more desirable to establish and maintain core cooling than to avoid venting. If the HPCS/RCIC and safety / relief valves are not operable (a very degraded and extremely unlikely case), another emergency means of venting the reactor must be used. It is emphasized, however, that such emergency venting would be in the interest of core cooling and therefore would be employed under emergency procedure guidelines. It is thus concluded that there is no reason to interfere with ECCS operation to avoid venting. It is further concluded that the emergency procedure guidelines, by correctly specifying operator actions for RCIC and SRV operation, also correctly specify operator actions to vent the reactor. In the event of HPCS failure and continued vessel pressurization, the effect of noncondensibles in the RCIC turbine steam was evaluated for three cases: O 1A.19-6 132A6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 lA.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 s_ Item II.B.1) (Cont'd) . L Response (Cont'd) i
}
- a. Continuous evolution of noncondensibles due to l
i radiolysis; .i b. Quasi-continuous evolution of noncondensibles due i to core heatup;
- c. The presence of a quantity of noncondensibles in the reactor at the time of RCIC startup.
Case a is a normal operating mode for RCIC and is of no concern. 4-l For Case b to exist, the core must be uncovered. Such '
! a condition requires multiple failures as shown in the t degraded cooling analyses. Core uncovery is prevented '
(or cladfing heatup into the rapid oxidation range is prevent ed) when only one ECC system is operating. For j a small pipe break or a loss of feedwater, which would i allow the reactor to remain at pressure, the HPCS and/or RCIC pumps would maintain inventory and taere would be no substantial hydrogen production. If neither ! HPCS-nor RCIC could maintain inventory, the reactor would be automatically or manually depressurized via safety / relief valves '(or via the break, for larger breaks). Low pressure water injection systems (LPCI or j LPCS) would then make up inventory. With the core l covered neither the rapid generation-of noncondensibles nor their accumulation would be possible. t O 1A.19-7 132A7
= GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 lA.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 Item II.B.1) (Cont'd) u Response (Cont'd) E l y The performance of RCIC under Case c is of concern only if there has been a very substantial producti,n of E hydrogen due to core uncovery and there is a need to l start the RCIC. This is extremely unlikely and an intolerable circumstance, because it could arise only if the core were allowed to remain uncovered for a long ' period with the reactor at high pressure. Automatic depressurization system operation and explicit operating instructions and the emergency procedures guidelines are intended to preclude this. If the level has fallen ' with the reactor at high pressure, the vessel would be E depressurized via the relief valves automatically or manually to permit low-pressure injection independent of RCIC performance. In the post-LOCA condition, it is possible to have noncondensible gases come out of solution while operating the residual heat removal (RHR) system in the shutdown cooling or steam condensing mode of operation. These gases would accumulate at the top of the RHR heat exchanger since this is a system high point and an area of relatively low flow. Gases trapped here will be r vented through a 3/4-inch vent line with two safety-related Class lE motor-operated valves operated from the control { room (as shown in Figure 5.4-12). As this vent line 5 and associated valves are part of the original design, they have also been considered in the design-basis . accident analysis contained elsewhere in this document. . To accommodate the continuous release of noncondensibles e 1A.19-8 132F3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.19 REACTOR COOLANT SYSTEM VENTS (NUREG-0737 Item II.B.1) (Cont'd) Response (Cont'd) from the RHR Heat Exchanger when employed in tge steam-condensing mode, these remote vent valves on the heat exchanger vent line are opened to discharge through a submerged line into the drywell portion of the suppression pool. Because the relief valves and RCIC will vent the reactor continuously, and because containment hydrogen calculations in normal safety analysis calculations assume continuous venting, no special analyses are required to demonstrate "that the direct venting of noncondensible gases with perhaps high hydrogen concentrations does not result in () violation of combustible gas concentration limits in containment." Conclusion and Comparison with Requirements The conclusions from this vent evaluation are as follows:
- a. Reactor vessel head vent valves exist to relieve head pressure (at shutdown) ta the drywell via remote operator action.
- b. The reactor vessel head is continuously swept to the main condenser and can be vented during operating ,
conditions via the SRV's to the supprerFion pool.
- c. Tae RCIC system provides an additional vent pathway to the suppression pool.
O 1A.19-9 4 132F4
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.19 REACTOR COOLANT SYSTEM VENTS (NOREG-0737 Item II.B.1) (Cont'd) Response (Cont'd)
- d. The size of the vents is not a critical issue because BWR SRV's have substantial capacity, exceeding the full power steaming rate of the nuclear boiler.
- e. The SRV's vent to the containment suppression pool, where discharged steam is condensed without causing a rapid containment pressure / temperature transient.
- f. The SRV's are not smaller than the NRC defined amall LOCA. Inadvertent actuation is a design-basis event and a demonstrated controllable transient.
- g. Inadvertent actuation is of course undesirable, but since the SRV's serve an important protective function, no steps such as removal of power during normal operation, should be taken to prevent inadvertent actuation.
- h. A dual inoication of SRV position (pressure and temperature) is provided in the control room.
- i. Each SRV is remotely operable from the control room.
- j. Each SRV is seismically and Class lE qualified.
- k. Block valves are not required, so block valve qualifications are not applicable.
lA.19-10 132A10
GESSAR II . ?2A7007 238 NUCLEAR ISLAND REV. 7 l 1A.23 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2) (NUREG-0737 Item II.D.1) NRC Position Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.
Response
A generic test program has been conducted through the BWR Owners Group to satisfy this requirement. The testing requirement to qualify SRVs for the " expected [~
'~'} operating conditions" associated with design-basis accidents and operational transients was determined through systematic analysis of those events as defined in Regulatory Guide 1.70, Revision 2. The conclusion from that evaluation was submitted to the NRC in September 1980 (Reference 25) in response to Item 2.1.2 of _
NUREG-0578 (Reference 12); the conclusion was that "there is no design-basis accident or transient which l requires safety, relief, or dual function SRV's to pass I i two-phase or liquid flow at high pressure." This l submittal, however, acknowledged the alternate shutdown ! cooling mode which is considered in the design analysis . , of plants and committed to testing SRV's with liquid l under low pressure conditions associ;1ted with this _ l avent. Additional justification for the conclusion that no high pressure liquid or two-phase discharge testing is (a~ ') s 1A.23-1 132F5 1
1 GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 l 1A.23 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2) (NUREG-0737 Item II.D.1) (Cont'd) l l Reeponse (Cont'd) l l reqttired was provided by the BWR Owners' Group to the NRC Staff during meetings on February 10, 1981 and March 10, 1981. A test plan which describes the test program for SRV testing for the alternate shutdown mode of cooling was l included in the September 1980 (Reference 25) submittal
~
to the NRC. The purpose of the test plan was two-fold:
- a. To demonstrate the capability of each type of SRV used in BWRs to operate satisfactorily under the f bounding case of expected water discharge release l I
of low-pressure water with resultant typical SRV j l discharge pipe loads on the SRV. . j b. To measure the SRV piping discharge loads during water discharge through these valves. The Dikkers 8 x 10 direct-acting SRV to be used in the 238 Nuclear Island was included in this test program. f The test program provides for consideration of remote manual initiation of the SRV's. Among other tests, the program involved the admission of slightly subcooled water at approximately 250 psig for fluid flow testing. This test followed a normal steam discharge test. This sequence of steam test - water test was repeated three times for each valve. I 1A.23-2 132F6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.23 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2) (NUREG-0737 Item II.D.1) (Cont'd) , Response (Cont'd) The acceptance criteria included proper opening on
- demand (inlet pressure at setpoint pressure); proper blowdown, i.e., SRV does not reclose except when inlet pressure drops below the setpoint minus the blowdown decrement; SRV to open properly on command for relief function; and pressure integrity of the valve body, connections, and piping is maintained at all times.
The generic test program has been completed and final - test results were' transmitted in a letter from T. J. () Dente (BWR Owners' Group) to D. G. Eisenhut (NRC), dated July 1, 1981. The results showed that all of the test criteria were met for all valves tested. As part of the response to this requirement, it was determined that the high drywell pressure inhibit of l the HPCS high level trip on the 238 Nuclear Island ' l should be removed, as it's removal decreases the probability of water entering the steam lines and the 1 i SRVs being subjected to high pressure water N1ow. The following paragraphs describe this modification as it relates to the 238 Nuclear Island design. i ! ' Current Design When a low reactor water level (L2) occurs in the l previously defined 238 NuclearLIsland design I (Figure 7.3-lb), tus L2 trip units will provide the one ( 1A.23-3 132F7
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.23 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2) (NUREG-0737 Item II.D.1) (Cont'd) Response (Cont'd) out of two, twice logic that initiates HPCS injection into the reactor vessel (5301 initiation signal). When the high drywell pressure inhibit signal is not present, this injection will continue until a high reactor water level (L8) is reached, and then the L8 trip unite will provide the two out of two, once logic which will terminate the HPCS injection automatically (5302 termination signal). If the high drywell pressure inhibit signal is present, the L8 trip is inhibited from initiating, and termination of the HPCS injection will not occur automatically. This high drywell pressure inhibit signal of the HPCS high water level trip can occur under LOCA conditions and non-LOCA conditions, such as a consequence of loss of drywell coolers, or a recirculation pump seal failure. Since the HPCS has the capability of raising the reactor vessel water level to an elevation that is above the main steamline nozzle, this inhibiting action can result in an increased probarility of high pressure liquid entering the main steamli) as, safety / relief valves (SRV) and being discharged via the SRV discharge lines to the suppression pool. Design Modification Figure 1A.23-1 shows the primary functional change which is being implemented to remove the high drywell pressure inhibit logic. The removal of this inhibit - 1A.23-4 132A15
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.23 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND () PRESSUPIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2) (NUREG-0737 Item II.D.1) (Cont'd) Response (Cont'd) logic assures that the L8 trip will function auto-matically to terminate the HPCS injection into the reactor vessel when the L8 trip is present. Separation of the cechanical divisions for initiation of the high water level t_-ip is required to ensure that a single failure of an instrument line does not cause an inadvertent trip at the HPCS system. In order to accommodate this change, the Level 8 trip unit for B21-N674G is reassigned to another division as shown in Figure 1A.23-2. The plant modifications for the HPCS trip logic described above will be reflected in Figure 7A.3-1 and Figure 5.1-3c following staff approval of this response. G' 1A.23-5/lA.23-6 132F8
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4
] 1A.24 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (NUREG-0737, Item II.D.3)
NRC Position Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.
Response
The 238 Nuclear Island will be equipped with a safety Relief Valve Open/ Closed Monitor (SRVOCM) in order to provide the operator with positive indication of valve position (closed or not closed). A positive system providing status information on the safety / relief O*s valves is required by NUREG-0737 Item II.D.3 to assist the plant operators during normal and abnormal operating conditions by providing the following information:
- 1. Positive indication of valve position including the " stuck-open" valve condition.
- 2. Positive identification of the specific valve or valves which are open.
- 3. Annunciation of activation of the Automatic Depressurization System (ADS) in the control room.
Providing prompt indication and annunciation of valve opening and identification of the valve, enables plant operators to initiate. appropriate action in a timely manner. Os 1A.24-1 16F5
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.24 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (NUREG-0737, Item II.D.3) (Cont'd) Response (Cont'd) As shown in Figure 1A.24-1, the Safety / Relief Valve Positive Open/ Closed Position Monitor System consists of three pressure switches connected by a hydraulic sensing line to the discharge piping of the Safety / Relief Valve. The output of each pressure switch is connected ~. to a circuit board assembly containing relays and electronic logic. The circuit board (s) are mounted in a metal NEMA-4 cabinet enclosure at a point outside the containment, remote from the pressure switch. An open S/RV pressurizes the discharge line and the hydraulic sensing line to the pressure switch, actuating the pressure switch. The electrical output of the pressure switch controls a sensor relay mounted on the circuit board at a point remote from the pressure switch. The relay contacts provide input to the annun-ciator in the control room, to the process computer and to an indicator light on a control room instrument panel. The pressure switches are designed for LOCA conditions. The pressure switches are activated by increasing pressure in the discharge line and will reset after the S/RV closes and pressure in the line decays. The pressure switches are mounted inside the containment and the circuit board and cabinet are mounted outside the containment in a convenient location. Each relay of the circuit board has isolated testability. O 1A.24-2 16El
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.24 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (NUREG-0737, item II.D.3) (Cont'd) Response (Cont'd) The system is rated for maximum vessel pressure. A leaking S/RV will not actuate the pressure switch, thus, the system detects only an appreciably open SRV to eliminate misleading indications to the plant operators. The pressure switches are commercial grade components designed for application inside the containment structure and capable of operation under LOCA environmental conditions. The pressure switches are designed for conditions in excess of the LOCA conditions of 340*F maximum temperature (6-hour period), 200*F continuous () operation at 100% relative humidity and 2.5G acceleration on all axes. These pressure switches are fabricated from materials which are compatible with the LOCA environmental conditions. The only nonmetallic materials are in the body of the microswitch which is designed for application in conditions which exceed the LOCA l environment. The circuit board assembly consists of a prewired board with F3nsor relays mounted. The circuit boards are mounted in a NEMA-4 type cabinet enclosure which has a maximum capacity of eight circuit board modules. The three-channel system provides a 2-out-of-3 logic. With this logic, failure of a single pressure switch will not compromise the reliability of the system and will not result in an erroneous valve position signal to the control room. O 1A.24-3 16F7
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.24 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (NUREG-0737, Item II.D.3) (Cont'd) Response (Cont'd) In the three-channel system, shown in Figure 1A.24-1, a single hydraulic sensing line from a SRV discharge pipe is connected by a manifold to three identical pressure switches in parallel. The pressure switches are elec-trically connected to the sensor relays in the circuit board assembly. The contacts of the relays are wired in a 2-out-of-3 logic which signals the annunciator, computer and indicator lights. These modifications to the 238 Nuclear Island will satisfy the requirements of NUREG-0737 Item II.D 3 and .. will be reflected in Figure 5.1-3a and section 5.2 following staff approval of this response. A diverse measurement for indication of SRV opening or long-term leakage is provided via temperature elements mounted in thermowells on each of the SRV blowdown pipes to the suppresion pool. These indications provide confirmation of the SRVOCM readouts. O 1A.24-4 16E2
4 GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.39 REACTOR VESSEL LEVEL INSTRUMENTATION O (NUREG-0737 Item II.K.1.23) NRC Position
- i i
! For boiling water reactors, describe all uses and types of reactor vessel level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that might give the operator the i same information on plant status. See Bulletin 79-08, 1
Item 4. I r
Response
The water level measurement for the 238 Nuclear Island is fully described in NEDO-24708A, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors." An outline of this description is provided in the following paragraphs. Figure 7.7-1 illustrates the reactor vessel elevations covered by each water-level range. The instruments that sense the water level are differential pressure devices calibrated to be accurate at a specific vessel pressure and liquid temperature condition. The following is a description of each water-level range. 1 I a. Shutdown water-level range: This. range is used to monitor the reactor water-level during-the shutdown condition when the reactor system is flooded for maintenance and head. removal. The water-level measurement design is the condensate reference chamber leg type that is.not compensated for p changes in density. The vessel temperature and 1 0 1A.39-1 i ! 132A24 l l
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.39 REACTOR VESSEL LEVEL INSTRUMENTATION (NUREG-0737 Item II.K.l.23) (Cont'd) Response (Cont'd) pressure conditions that are used for the calibra-tion are O psig afd 120*F water in the vessel. The two vessel instrument penetrations elevations used for this water-level measurement are located at the top of the RPV head and the instrument tap just below the bottom of the dryer skirt.
- b. Upset water-level range: This range is used to monitor the reactor water when the level of the water goes off the narrow-range scale on the high side. The design and vessel tap location are the same as outlined above. The vessel pressure and temperature conditions for accurate indication are at the normal operating points.
- c. Narrow water-level range: This range uses for its RPV taps the elevation near the top of the dryer skirt and the tap at an elevation near the bottom of the dryer skirt. The instruments are calibrated to be accurate at the normal operating points.
The water-level measurement design is the condensate reference chamber type, is not density compensated, and uses differential pressure devices as its primary elements. The feedwater control system uses this range for its water-level control and } indication inputs. O 1A.39-2 132F9
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.58 SEPARATION OF HIGH-PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEM INITIATION LEVELS -- ANALYSIS AND IMPLEMENTATION (NUREG-0737 Item II.K.3.13) (Cont'd) Response (Cont'd) plant experience was evaluated to estimate the frequency of occurrence of HPCS* and RCIC initiations. Based on this evaluation, it was concluded that the current design is satisfactory, and a significant reduction in thermal cycles is not achievable or necessary. Evaluation of Proposed Auto-Restart of RCIC The BWR Owners' Group sponsored a program to evaluate this concern and develop an appropriate modification. The results of this program were submitted to the NRC via a letter from D. B. Waters, Chairman of BWR Owners' Group, to D. G. Eisenhut, Director of NRC, dated
~
December 29, 1980 (Reference 20). These results conclude that automatic restart of the RCIC would contribute to improved system reliability and that it could be accomplished without adverse effects on system function and plant safety. Therefore, the 238 Nuclear Island design will be modified to allow automatic restart of l the RCIC system following its trip on high RPV water ! level. . l
*The HPCS system replaces the HPCI system in the 238 Nuclear j Island. The above referenced BWR Owners' Group analysis addresses the use of both systems.
i l
, lA.58-3 l
132F10 l
i GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 I 1A.58 SEPARATION OF HIGH-PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEM INITIATION l LEVELS -- ANALYSIS AND IMPLEMENTATION (NUREG-0737 Item II.K.3.13) (Cont'd) Response (Cont'd) The plant modifications to allow automatic restart of the RCIC system following its trip on high RPV water level will be reflected in Subsection 5.4.6 following staff approval of this response. A technical description of this modification is included in Section 15D.2.1.3.1. O 1A.58-4 O 132A31
i GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.61 REPORT ON OUTAGES OF EMERGENCY CORE-COOLING O# SYSTEMS LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES (NUREG-0737 Item II.K.3.17) NRC Position Several components of the emergency core-cooling (ECC) systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours for one diesel-generator; 14 days for the HPCI system). In addition, there are no cumulative outage time limitations for ECC systems. Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include the causes of the outages (i.e., controller failure, spurious isolation).
; Response The response to this requirement will be supplied by
~'
licensees of plants with sufficient operating time ( ~ to provide useful data. This requirement does not apply to NTOL plants or any future plants. l 1A.61-1/lA.61-2 125L5
3 % m (J s J ICOLO RE AC tog CON isFA ROG M WATE R LEVEL I N DIC A T ION AND T Rip LEVE(5 VESSEL) INST RUM E N T(53 SEE NOTE 3 VESSEL p RE6ERENCE tNCHE S DE SCR PTION PRoviD'NG LE SAFEGUARDS lE EDWATE R
; gg' g, F m 20~E .m RA~a oppo. R.~x UPSE 7 5~TDO.h LW W615 LktPW N623A fs C 33 fC34 LR N608 C 3 3rC 34 LiR605 La R6tO Li 9604 C31/C 34 Li R606ABC LR R60 8 4000-180.O '
TOPO, 4A3 ,L.NcE eSe+ STE A M LINE I 644.5' NO22 L E J iNSTRyutN T L eNg i f 06.0 " NOZZLE j ' Trip RCIC TUR&NE & MPCS
- 60.0" E*10 6C .0 "
H IN ACTION VALVE CLOSURE SsGNA L , CLOSE MAIN TURBihE StP VALVES ~ TABLE U REF 2 8 *iS O5' 55 05*
~ ~ ~ ~ ~ ~ ' '----~~~~
TR P FEED PUMP 5 AND REF 1 CONDENSAYE ECOs*ER
,PJMPS. SCRAM. .
HIGH LEVEL AL ARM REF 1 ------ - - - - - - - - - - - -
,__, NCn9M AL WATE R LEVEL RE F t ,
LOW L E VE L AL AR M REF 1 - - - - - - .- - - - - - - - - - - - - - RUN RECIRC FLOWBACM R W
'5 CRAM & CONTRIBUTE TO ~ TABLE D REF 2 3
~ - - - " " '------~~
10.3S" tOJS' i AV O DEPRESSURil.A TION. REF i RUN RECIRC FLOW BACK H CLOSE RHR $HUTDOWN Q M J AAYER LE sE L ISOL valves Q trj e . N 57 RUMENT ZERO
$3 3 O* O- 0- 0- O '- 0- UU Q NOM OF De rE R
~ ~ ~ ~ " ~~--~~~
MU su aR T + tS'. kk W IN*,T RUMENT LINE I S t 8.O ' INITIATE RCIC & HPCS NN NO22 E I CLOSE T1hMt.RY Sv5T E MS iSOL valve S E XCEPT HH kHR SHU TDOW N t%CL 2 . 36.5 5 H
- , VALVES & M51v'S - TABLE D REF 2 - - - - - - -
M Trip RECIRC PUMP 5 START
-w p
. , Div 3 STAND-BY Df 5EL. g ,
-ll kl. 6,, g INITIATE RHR & L PC5.
i CONTRIBUTE TO AUTO t . e4s 2 5-K DE PRE 55UR12 A TION. + TA BLE D RE F 2 - - - - - - START Dv t & Dv 2 - 16 0. 0 " STAND-BY DESELS. CLOSE M5sv'S
- M. X TOP er AC 'i /E F.,EL1 363 5' ~8""
& FUEL ZONE *ATER ~ ~ ~ ~ ~
LEVEL ZERO l e MODIFICATION 1 36 4.O ' sNSTRUMENT LINE 386. l f" NOZZLE f JE T PUMP D# FUSED I M L TAP I JE T PUMP INSTRU- 1 156 5' WENT NO22LE j
# FUNC TWON t~ 4 FEEDWATER CONTROL SYS t REF t) FOR LO55 OF ONE F EED PUMP M NN to >
Figure 1A.67-1. Planned Modification for Common Water .# o C Level Reference (Figure 5.2-11) y 3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 O
/
-120IN.
4
-160 IN.
TAF AT O PSIG
-200 IN.
=e e
-240 IN.
-280 IN.
-320IN.
/
FUEL ZONE (Ll R610) i Figure 1A.67-2. Fuel Zone Common Reference Zero lA.67-4
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.68 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEM VALVES (NUREG-0737 Item II.K.3.28) NRC Position Safety analysis reports claim that air or nitrogen accumulators for the automatic depressurization system (ADS) valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated that the emergency core cooling (ECC) systems are designed to withstand a hostile environment and still perform their function for 100 days following an accident. Licensee should verify that the accumulators on the ADS valves meet these requirements, even considering normal leakage. If this cannot be demonstrated, the Licensee must show that the accumulator "N design is still acceptable. (d
Response
The accumulators for the ADS valves are sized to provide two operating cycles at 70% of drywell design pressure. This cyclic capability is validated during preoperational testing at the station. The accumulators are safety grade ASME Section III componente. The 100-day, post-accident functional operability requirement is met through conservative-design and redundancy; eight ADS valves are provided with code-qualified accumulators and seismic Category I piping within primary containment. Two redundant 7-day supplies of bottled air are available for long-term usage with , 1 replacement capability being provided for the remainder ; N of the postulated accident to assure system functional s_sl l lA.68-1 16C8
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.68 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEM VALVES (NUREG-0737 Item II.K.3.28) (Cont'd) Response (Cont'd) operability. A mtximu.. of three ADS valves need function _ to meet short-term demands and the functional operability of only one ADS valve will fulfill longer term needs. Each accumulator is instrumented to provide the reactor operator with indication of the failure of any of the redundant systems under hostile environmental condition. The BWR Owners' Group sponsored an evaluation of the adequacy of the ADS configurations. Evaluation results are summarized in the following paragraph. The accumulators are designed to provide two ADS actua-tions at 70% of drywell design pressure, which is equivalent to 4 to 5 actuations at atmospheric pressure. The ADS valves are designed to operate at 70% of drywell design pressure because that is the maximum pressure for which rapid reactor depressurization through the ADS valves is required. The greater drywell design pressures are associated only with the short duration primary system blowdown in the drywell immediately following a large pipe rupture for which ADS operation is not required. For large breaks which result in higher drywell pressure, sufficient reactor depressuri-zation occurs due to the break to preclude the need for ADS. One ADS actuation at 70% of drywell design pressure is sufficient to depressurize the reactor and allow inventory makeup by the low pressure ECC systems. However, for conservatism, the accumulators are sized to allow 2 actuations at 70% of drywell design pressure. See Subsection 6.8.1 for a description of the ADS air supply. 1A.68-2 16E3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.71 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE O FAILURE TO VERIFY NO FUEL FAILURE (NUREG-0737 . Item II.K.3.44) . NRC Position For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which result from a stuck-open relief valve should be included in this category.
Response
The BWR Owners' Group sponsored an evaluation of the worst anticipated transient with the worst single Os failure. These results were submitted to the NRC via a letter from D. B. Waters, Chairman BWR Owners Group, to D. G. Eisenhut, NRC, dated December 29, 1980. (Reference 20) . A letter (Reference.26) from D. G. . Eisenhut (NRC) to D. B. Waters (BWR Owner's Group) transmitted the NRC evaluation of this item. The staff found that the report was acceptable for referencing by individual licensee / applicants. A summary of the results of the analysis follows. The anticipated transients in NRC Regulatory Guide 1.70, Revision 3 were reviewed for all BWR product lines from the BWR/2 through BWR/6 from a core cooling viewpoint. l l The loss of feedwater event was identified to be the y most limiting transient which would challenge core I cooling. The 238 Nuclear Island is designed so that the HPCS, RCIC or ADS with subsequent low pressure-1A.71-1 16E4
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.71 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO VERIFY NO FUEL FAILURE (NUREG-0737 Item II.K.3.44) (Cont'd) Response (Cont'd) makeup is each independently capable of maintaining the water level above the top of the active fuel given a loss of feedwater. The detailed analysis in Reference 20 shows that even with the worst single failure in com-bination with the worst transient the core remains covered. Furthermore, even with degraded conditions involving one SORV in addition to the worst transient with the worst single failure, these studies show that the core remains coyered during the whole course of the transient either due to RCIC operation or due to automatic depressurization via the ADS or manual depressurization by the operator so that low pressure inventory makeup can be used. It is concluded that for anticipated transients combined with the worst single failure, the core remains covered. Additionally, it is concluded that for severely degraded j transients beyond the design basis where it is assumed that a SRV sticks open and an additional failure occurs, l the core remains covered with proper operator action. l O 1A.71-2 16C14
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 rs 1A.74 UPGRADE EMERGENCY PREPAREDNESS (NUREG-0737 Item III.A.l.1) NRC Position
- Comply with Appendix E, " Emergency Facilities" to 10 CFR Part 50, Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants," and for the offsite plans, meet essential elements of NUREG-75/111 or have a favorable finding from FEMA.
i Response
- The response to this requirement will be supplied by the applicant.
Section 1A.8 provides additional information on procedures for transients and accidents. O _ i I
*This position statement is repeated from Reference 22 since it was not given in detail in NUREG 0737.
lA.74-1/lA.74-2 l 132Fll
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4
- f. lA.80 REFERENCES
- 1. U. S. Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result of the TMI-2 Accident,"
USNRC report NUREG-0660, Vols. 1 and 2, May, 1980.
- 2. U. S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," USNRC Report NUREG-0737, November, 1980.
- 3. Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees and Applicants,
Subject:
Interim Criteria for Shift Staffing, dated July 31, 1980.
- 4. Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants,
Subject:
Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated September 13, Os 1979.
- 5. Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,
Subject:
FolloNup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated September 27, 1979.
- 6. Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,
Subject:
Emergency Planning, dated October 10, 1979. l
- 7. Letter from H. R. Denton, NRC, to All Operating i Nuclear Power Plants,
Subject:
Discussion of t Lessons Learned Short-Term Requirements, dated' l October 30, 1979. ( O 1A.80-1 l 16C18
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.80 REFERENCES (Cont'd)
- 8. Letter from D. B. Vassallo, NRC, to All Pending Operating License Applications,
Subject:
Discussion of Lessons Learned Short-Term Requirements, dated November 9, 1979.
- 9. Letter from D. G. Eisenhut, NRC, to S. T. Rogers, BWR Owners' Group,
Subject:
Emergency Procedure Guidelines, dated October 21, 1980.
- 10. Letter from D. B. Vassallo, NRC, to All Pending Construction Plant Applicants.
Subject:
Discussion of Lessons Learned Short-Term Requirements, dated November 9, 1979.
- 11. U. S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Final Report," USNRC Report NUREG-0585, October 1979.
I
- 12. U. S. Naclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," USNRC Report NUREG-0578, July, 1979.
- 13. U. S. Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result of the TMI-2 Accident," .
USNRC Report NUREG-0660, Appendix C, Table C.1, . item 5.
- 14. U. S. Nuclear Regulatory Commission, Office of Inspection and Enforcement Region III, " Nuclear Incident at Three Mile Island-Supplement," IE Bulletin No. 79-05B, April 1979.
O 1A.80-2 16E5
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.80 REFERENCES (Cont'd) O
- 15. U. S. Nuclear Regulatory Commission, "TMI-Related Requirements for New Operating Licenses," USNRC Report NUREG-0694, June 1980.
- 16. Letter from D. F. Ross, NRC, to All B&W Operating Plants (except TMI-1 and -2),
Subject:
Identification and Resolution of Long-Term Generic Issues Related to the Commission Orders of May 1979, dated August 21, 1979.
- 17. Reports of the Bulletins and Orders Task Force of the NRC Office of Nuclear Reactor Regulation:
- a. U. S. Nuclear Regulatory Commission, " Generic Evaluation of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in Westinghouse O Designed Operating Plants," USNRC Report NUREG-0611, January 1980.
- b. U. S. Nuclear Regulatory Commission, " Staff Report of the Generic Assessment of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in Boiling Water Reactors Designed by the General Electric Company," USNRC Report NUREG-0626, January 1980.
- 18. U. S. Nuclear Regulatory Commission, " Generic Assessment of Delayed Reactor Coolant Pump Trip During Small-Break Loss-of-Coolant Accidents in Pressurized Water Reactors," USNRC Report NUREG-0623, November 1979.
O 1A.80-3 16C20
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.80 REFERENCES (Cont'd) l
- 19. Letter from D. B. Waters, Chairman, BWR Owners' Group, to D. G. Eisenhut, NRC, dated March 31, 1981,
Subject:
BWR Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18.
- 20. Letter from D. B. Waters, Chairman, BWR Owners' Group, to USNRC dated December 29, 1980,
Subject:
BWR Owners Group Evaluation of NUREG-0737 Require-ments.
- 21. Letter from D. B. Waters, Chairman BWR Owners' Group, to USNRC, dated May 22, 1981, Ltr No.
BWROG-8142,
Subject:
BWR Owners' Group Evaluation of NUREG-0737, Item II.K.3.25, "Effect ef Loss of Alternating Current Power on Pump Seals."
- 22. U. S. Nuclear Regulatory Commission (FEMA-REP-1),
" Criteria for Preparation and Evaluation of Radio-logical Eme:::gency Response Plans and Preparedness in Support of Nuclear Power Plants," USNRC Report NUREG-0654, January 1980.
- 23. Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,
Subject:
Clarification of NRC Site Requirements for Emergency Response Facilities at Each Site, dated April 25, 1980.
- 24. Letter from R. H. Buchholz, GE, to D. F. Ross, NRC,
Subject:
Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, November 30, 1979, MFN-290-79. . lA.80-4 16E6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1A.80 REFERENCES (Cont'd)
- 25. Letter to R. H. Vollmer (NRC) from D. B. Waters (BWR Owners Group), "NUREG-0578 Requirement 2.1.2 -
Performance Testing of BWR and PWR Relief and Safety Valves", September 17, 1980.
- 26. Letter from D. G. Eisenhut (NRC) to all Applicants /
Licensees Referencing BWR etc.,
Subject:
NUREG-0737, Item II.K.3.44 - Evaluation of Anticipated Transients Combined with Single Failure (Generic Letter No. 81-32), August 7, 1982. O t O 1A.80-5 16E7 i
GESSAli II 22A7007 238 NUCLEAR ISLAND REV. 7 1AA.2
SUMMARY
OF SE: 'LDIhG DESIGe' EEVIEW (Continued) rooms and pumps and valves per Table lAA-1. All vital equipment will be environmettally qualified. It is also shown that this exposutb onvelope is not time dependent after about 100 days. c) It is not necessary for operating personnel to have
! access to any place other than the Control Room and three manual valves in the Auxiliary and Fuel Buildings to operate the equipment of interest during the 100 day l
period. The manual valves are for essential service water supply (one in each division) to the hydrogen mixing blowers of the Combustible Gas Control system and a Drywell Bleed-off Vent System valve. These _. valves are considered accessible on a controlled exposure basis. Direct shine from the containment is , less than one R/hr within four hours post-accident. d) The control room is designed to be accessible post-accident. e) Access to radwaste is not required, but the Radwaste Building is accessible since primary containment sump discharges are isolated and secondary containment sump pump power is shed at the onset of the accident. Thus, ( fission products are not transported to radwaste. The hydrogen control system is operated from the Control Room; the 238 Nuclear Island does not have a containment isolation reset control area or a manual ECCS alignment l area. These functions are provided in the control Room. O lAA-5 105G1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.2
SUMMARY
OF SHIELDING DESIGN REVIEW (Continued) f) Within days to a month following an accident, access would be possible to electrical equipment rooms con-taining motor control centers and corridors in the Auxiliary Building and to various areas in the Fuel Building. This is based on :adiation shine from the ECCS rooms and primary conttinment; there is no airborne radiation source ir the electrical equipment rooms and ECCS corridor are. While not necessary to maintain safe shutdown, such access can be useful in extending system functionality and in plant recovery. g) The emergency power supplies (diesel generators) are accessible within about 100 hours post-accident per Figure 1AA-1. However, access is not necessary since the equipment is environmentally qualified. O O 1AA-6 105CS
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.3 CONTAINMENT DESCRIPTION AND DEFINITION OF TERMS 1AA.3.1 Description of Primary / Secondary Containment The 238 Nuclear Island includes many features to assure that personnel occupancy is not unduly limited and safety equipment, is not degraded by post-accident radiation fields or during other operating periods. While these features are described in other sections, a brief review in the present context is provided here for emphasis. The configuration of the drywell and primary containment and the suppression pool maximizes the scrubbing action of fission products by the suppression pool. The particulate and halogen content of the primary containment atmosphere following an accident is thereby substantially reduced compared to the Regulatory Guide 1.3 source terms. The calculations for this design review were madu prior to O'N General Electric Company's suppression pool scrubbing i tests (Subsection 15D.2.2) so these calculations do not reflect the substantial decontamination factors found in the tests. The calculations therefore have extensively higher primary containment (and secondary containment) airborne radioactivity concentrations than would be the real case. The secondary containment consists of the Shield Building annulus, the Fuel Building and the ECCS rooms of the- Auxiliary Building. Primary containment leakaga is limited to less l than one percent of the primary containment volume per day l by the construction and by the Isolation valve Leakage Control Systems, Subsection 6.5.3.3, which seal penetration isolation valve leak paths. These control systems are
; unique to the 238 Nuclear Island design. Of this one percent, leakage to the Fuel Building and ECCS portion
[ 105C6 1AA-7 _ _ ~ -- . - _ _ . _ - _ - _ _ _ _ _ _ - - _ _ _ _ _
}
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AA.3.1 Description of Primary / Secondary Containment (Continued) of the Auxili.iry Building is less than 8%, i.e., 0.0008 of the primary containment volume per day. Entry of fission products and their accumulation in the Fuel Building is limited. Thi sottree term for the airborne level in these areas is thus mil; nal. The Standby Gas Treatment System (SGTS) operates automa-tically from the beginning of an accident to control the .. secondary containment pressure to (-)l/4" w.g. The Shield Building acts as a mixing chamber to dilute any primary containment leakage before processing by SGTS and discharge to the environment. Discharge of radioactivity is thus controlled and reduced. Radioactivity content of secondary containment atmosphere is reduced with time by SGTS treatment as well as by decay. (However, prior removal of halogens by scrubbing in the suppression pool offsets the degree of this treatment). Each ECCS pump and supporting equipment is located in an individual shielded, watertight compartment. Spread of radioactivity among compartments is thus limited. Radiation to the corridors and other areas of the Auxiliary Building is limited to shine through the walls; there is no airborne radiation in these other areas. These areas, outside of ECCS rooms, contain plant electrical control equipment, portions of leakage control systems and HVAC systems. When these become accessible (see Table lAA-7), any component failures can be repaired thereby improving systems avail-ability. lAA-8 105G2 i
, - . . - ~ - _ _ _ - I GESSAR II 22A7007 1 238 NUCLEAR ISLAND REV. 7 , () 1AA.3.2 Vital Area 1 A vital area is any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident. Areas which must be considered as vital after an accident , are the Control Room, Technical Support Center, Sampling Station and Sample Analysis area. The vital areas also include consideration (in accord with NUREG-0737, II.B.2) of the post-LOCA hydrogen control system, containment isolation reset control area, manual ECCS alignment area, motor control i
- centers, instrument panels, emergency power supplies, Security
! Center and radwaste control panels. Other areas specific to the 238 Nuclear Island to be considered are those for the i t 1 ADS pneumatic air supply and the auxiliary systems necessary ' ] for the operation of the ECCS systems, i.e., power, cooling I () water, and air cooling. Those vital areas which are plant unique as to location, i.e. Technical Support Center and Security Center, are normally areas of mild environment allowing unlimited access and therefore, are not reviewed
~
for access. lAA.3.3 Post Accident Operation Post-accident operations are those necessary to 1) maintain the reactor in a safe shutdown condition, 2) maintain adequate core cooling, 3) assure containment integrity and 4) control of radioactive ventilation releases within'10CFR100 guidelines. 1 Many of the safety related systems are required for reactor protection or to achieve a safe shutdown condition. However, they aren't necessarily needed once a safe shutdown condition is achieved. Thus, the systems considered herein are only-() the Engineered Safety Features (ESF) (see Chapter 6) used to maintain the plant in a safe shutdown condition. lAA-9 i 105G3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.3.3 Post Accident Operation (Continued) For purposes of this review the plant is assumed to remain in the safe shutdown condition. The basis for this position is that the foundation of plant safety is the provision of sufficient redundancy of systems and logic to assure that the plant is shutdown and that adequate cora cooling is maintained. Necessary shutdown and post-accident operations are performed from the Control Room, except for the several manual valves noted carlier. O O 1AA-10 105C9
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AA.4 DESIGN REVIEW BASES 1AA.4.1 Radioactive Source Term The radioactive source term used is equivalent to the source terms recommended in Regulatory Guides 1.3 and 1.7 and Standard Review Plan 15.6.5 with appropriate decay times. Depressurized coolant is assumed to contain no noble gas. There is no leakage outside of secondary containment other than via SGTS. Dose rates for areas requiring continuous occupancy may be averaged over 30 days to achieve the desired <l5 mrem / hour. Design dose rates for personnel in a vital area are such that the guidelines of General Design Criteria (GDC) 19 (i.e., <5 Rem whole body or its equivalent to any part of () the body) are not exceeded for the duration of the accident, based upon expected occupancy. lAA.4.2 Safety-Related Equipment Requirinq Qualification ] The safety-related equipment requiring review for qual- ] ification is only that necessary for post-accident opera-tions and for providing information for assuring post-accident control. t lAA.4.3 Qualification Duration of Safety-Related Equipment l l In 10CFR50 the long-term cooling capability is given as follows: " ... decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core." A 100 day period has been selected as a sufficient extended period for equipment qualification permitting site and facility response to terminate the event. (} 1AA-11 105G4
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.4.4 Availability of Off-Site Power Since the availability of off-site power is not influenced by plant radiation levels, recovery of off-site power can be achieved without consideration of plant accident conditions. Therefore, even though loss of off-site power may be assumed as occurring coincident with the beginning of the accident sequence, continued absence of off-site power for the accident duration is not realistic. While restoration of off-site power is not a necessary condition for maintaining core cooling, its availability can permit operation of other plant sytems which would not otherwise be permitted by emergency power restrictions, e.g. operation of the Pneumatic Air System, non-safety related HVAC systems and other systems useful to plant cleanup and recovery. Based on Table A.6-2 of Section 15D.3, the probability for off-site power recovery is 0.999 in 15 hours. This is conservative since the longest time for restoration of off-site power was six hours for the Pennsylvania-New Jersey-Maryland interconnection, the grid used as a basis for the probabilistic risk assessment presented in Section 15D.3. 1AA.4.5 Accidents Used as the Basis for the Specified Radioactivity Release Table 15.0-3 summarizes the various design basis accidents and associated potential for fuel rod failure. This chapter also provides the accident parameters. Of those accidents only the DBA-LOCA may produce 100% failed fuel rods under NRC worst-case assumptions. The rod drop accident and fuel handling accident are the only other accidents postulated as leading to failed fuel rods with the potential consequence of radioactivity releases. lAA-12 105C11
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 ("'s lAA.4.6 Environiaental Qualification Conditions (Continued) , Radiation sources in the secondary containment (especially the ECCS rooms of the Auxiliary Building) are the same as the Table 1AA-2 design basis values for water sources. For airborne radiation sources the plant design basis of Table 1AA-2 for air is used except the primary containment leakage rate to the equipment areas of secondary containment (8% of 1%) is used. Conservatively, the entire 0.08% primary containment leakage is assumed to occur in each of the individual secondary containment compartments. This leakage is limited by the Fission Product Control Systems (Sub-section 6.5.3.3). As previously noted, no credit has been taken for the radio-halogen scrubbing which is an inherent feature of the BWR. 1AA-15/lAA-16 O 105G5
4 GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.5 RESULTS OF THE REVIEW lAA.S.1 Systems Required Post-Accident This section establishes the various systems which are required to function following an accident along with their locations. In the following sections the expected environ-mental conditions and access and control needs are established. for the required post accident systems. 1AA.5.1.1 Necessary Post-Accident Functions and Systems Following an accident and assuming that immediate plan-recovery is not possible, the following functions
- are necessary:
a) Reactivity control () b) c) Reactor core cooling Reactor coolant system integrity d) Primary reactor containment integrity, and e) Radioactive effluent control Reactivity control is a short-term function and is achiev when the reactor is shutdown. The remaining functions are achieved in the longer term post-accident period by use of: a) The Emergency Core Cooling System (ECCS)'and-their auxiliaries (for reactor core cooling), j t
- ANSI /ANS 4.5 Criteria for Accident Monitoring Functions in Light Water Cooled Reactors i
1AA-17 105C15
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AA.5.1.1 Necessary Post-Accident Functions and Systems (Continued) b) The Fission Product Removal and Control Systems and auxiliaries (for containment integrity and radioactive effluent control), c) The Combustible Gas Control System (CGCS) and auxiliaries (for reactor coolant system and primary containment integrity), and d) Instrumentation and controls associated with the post-accident monitoring and functioning of the above systems and associated Habitability Systems. 1AA.5.1.2 Emergency Core Cooling Systems and Auxiliaries Table 1AA-3 shows various systems related to cooling the fuel under post-accident conditions. The table has two purposes: a) to show what major cooling equipment and systems are required to function simultaneously and thereby define the systems for review, and b) to show the equipment locations. l This table shows for example that a diesel generator, ECCS equipment and equipment coolers in a ECCS room and essential service water in the same division must all perform together to provide an ECCS function. 1 I l As indicated by Table 1AA-3 and Subsection 6.3.1.1.2, it is required that any two of the three combinations tabulated under divisions 1, 2, 3 plus ADS are necessary to achieve i safe shutdown within the single failure criterion and loss 1AA-18 105G6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 Cs lAA.S.l.2 Emergency Core Cooling Systems and Auxiliaries V (Continued) of off-site power. Under these assumptions at least one RHR heat exchanger is available for cooling purposes. The cooling function can also satisfy the containment cooling function in that by cooling suppression pool water, which is the source of water flowing to the reactor, the containment source of heat is also removed. The diesel generator, electrical switchgear and essential service water system of the same divisions, will also be needed. As a minimum, this combination must last until other actions can be taken (e.g., for 100 days). However, see Section 4.4 regarding the needs for diesel generators. The fuel pool cooling function is also included in Table 1AA-3 on the basis that a recently unloaded fuel batch could require continued cooling during the post-accident period. The FPCCS equipment is environmentally qualified'so access is not required and redundancy is included in system components. The Automatic Depressurization System (ADS) pneumatic air supply is included in the tabulation since a postulated non-break or small break accident could require continued need for the depressurization function until the RHR system is placed in the shutdown reactor cooling mode. In the case of a non-break or a small break accident, the majority of the fission products would be released via the safety relief valves to the suppression pool and hence to the containment rather than direct mixing through the horizontal vents as would occur following a DBA-LOCA. In either case the distri-bution of fission products is assumed to be the same as for the DBA-LOCA even though realistically a significant portion i O 1AA-19 105C17
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 i l 1 1AA.5.1.2 Emergency Core Cooling Systems and Auxiliaries (Continued) 1 of halogens and solid fission products would be retained i l in the reactor pressure vessel. Thus, the results as they apply to the ADS are very conservative. Table 1AA-4 supplements Table 1AA-3 by showing the location of selected associated valves and instrument transmitters. These do not represent all of this type of equipment which is environmentally qualified, safety-related, or included in the systems of Tables 13.11-2 through 13.11-7. It does however, represent components which are needed to operate, generally during the beginning of the accident. For example, most ECCS system valves are normally open, and only a pump discharge valve needs to open to direct water to the reactor. Similarly, the instrument transmitters shown are those which would provide information on long term system performance post-accident. Control Room instrumentation is not listed since it is all in an accessible area where no irradiation degradation would be expected. Passive elements such as thermocouples and flow sensors are not listed although they are environmentally qualified. The components listed under B21-Main Steam are those for ADS function or monitoring reactor vessel level. Suppression pool level is included with the HPCS instrumentation. 1AA.5.1.3 Fission Product Removal and Control Systems and Auxiliaries The systems and equipment of interest are shown in Table 1AA-5 hl and are described in Section 6.5 (except as noted). Included are: lAA-20 105G7
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 Discussion of Conservatism in Calculations O 1AA.S.2.2 (Continued) The electrical equipment rooms-in the Auxiliary Building are outside the secondary containment barrier and so are not exposed to airborne radiation. They are located two floors I above the HPCS or LPCS rooms so dose rates calculated at ] three feet above the HPCS or LPCS cover slabs are greater than expected at the equipment rooms and the electrical equipment exposures have been reduced accordingly. Although access is not required during the post-accident period, access is possible for short times after 5 days (2 R/hr) and for progressively longer times after 30 days (400 mR/hr) and 100 days (70 mR/hr). i The above discussion shows that the calculated exposures are high compared to what would actually be the case. The cal-
, fg culations would envelope the actual expected conditions but k/ still be within the environmental requirements. As shown ; later, the calculated conditions are within the environmental qualification conditions.
lAA.S.2.3 Discussion of Fuel Building Access Needs Dose rates and exposures in the Fuel Building are of post-accident interest because the following are located there: i a) SGTS equipment b) ADS air bottles and air compressors and receivers c) Fuel Pool Cooling and Cleanup Systems d) The isolation valve leakage control system mechanical aquipment O
.lAA-25 105C23
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AAS.2.3 Discussion of Fuel Building Access Needs (Continued) Equipment which is qualified to the environmental require- 1 ments will operate through the post-accident period so access is not required for that reason. While the Water Positive Leakage Control System water supply is designed for a minimum of 30 days without make-up, additional water can be supplied by the Essential Service Water system. The safety related air compressors for the isolation valve leakage control systems are redundant. Thus, these systems will continue to provide isolation valve seal leakage control. The air supply to the Automatic Depressurization System " (ADS) is adequate for seven days in which case replacement of the air bottles could be necessary. In the event of a DBA-LOCA this would not be required. In the event of a non-break or small break LOCA it may be necessary, but in that case the probability of a concurrent fission product release as postulated for the DBA-LOCA is small. Thus, access to the Fuel Building for bottle replacement is expected to be possible. On the other hand restoration of off-site power is highly probable within 15 hours after the accident. This would permit operation of the non-safety grade pneumatic air systen compressors which also supply air to the ADS. These compressor systems are redundant and are in a mild environment; thus access to replace the ADS air bottles is not considered necessary under post-LOCA conditions. . O 1AA-26 105G8
GESSAR II .2A7007 238 NUCLEAR ISLAND REV. 7 1AA.6 CONCLUSIONS OF THE REVIEW Table 1AA-7 shows that the dose rates in the ECCS rooms will not permit occupancy during the post-accident period of 100 days. However, since the equipment in these rooms will be qualified for operation throughout this period, occupancy is not expected, nor required. At 7 R/hr after 30 days and 2 R/hr after 100 days the ECCS corridor could be visited for a short time; however, there is no need since equipment located there will be environ-mentally qualified. Only corridor locations adjacent to operating ECCS rooms will have this conservative dose rate. , , At 400 mR/hr after 30 days the electrical equipment rooms can be considered accessible; this drops to 70 mR/hr at 100 days. However, access is not required for continued ECCS O g ,/ operation. Again, only equipment above operating ECCS rooms will have these conservative dose rates. Auxiliary Building self-contained A/C units are environmentally qualified and will provide cooling air to the electrical equipment areas. The Fuel Building radiation results primarily from airborne radiation due to assumed leakage from the primary containment except for the SGTS and FPCCS areas. In these areas the i exposure is principally due to the radioactive material i collected on the SGTS filters and circulating in the FPCCS. At 100 days the general Fuel Building dose rate is 20 mR/hr, so at some time after 30 days access is possible for workable periods within the NUREG-0737, II.B.2 exposure guidelines. Figure 1AA-1 shows the dose rates and integrated dose in the l Diesel Generator Building. The Diesel Generator Building is accessible within about 100 hours. However, the equipment
- () is environmentally qualified so access is not required.
lAA-27 105G9
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AA.6 CONCLUSIONS OF THE REVIEW (Continued) From an equipment viewpoint, Figure 1AA-2 shows the dose rate and integrated exposure for the ECCS rooms as a function of time. Three curves of integrated dose are shown for three conditions: 1) average piping dose is derived from the dose rate curve of the figure, 2) the integrated dose curve is based on the peak piping dose rate and gives the highest calculated exposure, 3) the third curve assumes no water leakage and reflects the fact that an operating pump room cannot also have significant leakage over a 100-day operating period because it would be flooded. All of the _ exposure curves show less integrated dose than the qualifica- . tion exposure. Therefore, equipment qualified to the requirements would operate through the specified post-accident period. Since the dose rate declines with time, the calculated integrated dose is always below the required dose regardless of time and does not change appreciably after 100 days. Thus, exposure would not be a factor limiting equipment performance. Table 1AA-9 compares post-accident and equipment qualification exposures in the areas of interest. The equipment which is qualified to the indicated values will operate through the specified post-accident periods. Since access is not required for equipment qualified to the environmental requirements, access is not required except in the control room. No changes are therefore necessary. This review has shown that the requirements of NUREG-0737, II.B.2 are satisfied. O 1AA-28 105G10
GESSAR II 22A7007
- 238 NUCLEAR ISLAND R3v. 7 ,
i TABLE lAA-1 ! REQUIRED ** ACCIDENT ENVIRONMENTAL CONDITIONS i Location
- Equipment Integrated Radiation, Rads a
Beta Gamma 8 CT-1 Some instrument 3.3x10 2.2x10 transmitters C AB-1 Elect. Switchgear (1x10 1.7X10 1 Remote Shutdown Panel Rooms 6 AB-2 LPCS, HECS, RHR "C" 5x10 5x10 (3) l Rooms 6 i AB-4 RHR Pump Rooms "A" & 5x10 5x10 (3) . ) "B" AB-6 Corridors outside - 1.7x10 i ECCS Rooms . 4 FB-1 Fuel Pool Pump Areas 4x10 2x10 (4) 4 FB-2 Operating Floor 4x10 2x10 4 FB-3 Below Operating Floor 4x10 2x10 3 3 ! CB-1 Control and Control <5x10 < 1x10 . Equipment Rooms 4 4 DG Bldg. Diesel Generator and 4.5x10 1.0x10 j Auxiliaries incl. HVAC l i *See Table 3.11-1 for locations- . ! ** Required is that accident exposure which, when added to the non-accident exposure, will be used for equipment qualification. , Note 1) Exposure is based upon a condition duration of 100 days of DBA-LOCA.
- 2) Small break accidents have the same or lesser specified
- exposures.
7
- 3) RHR and LPCS pumps.and valves = 1.4x10 Rads, gamma.
- 4) SGTS cubical bar 1.4 x 10 Rads, gamma total. -]
n U lAA-29 ,
GESSAR II 22A7007 238 NUCLEAR ISLAhD Rev. 7 , TABLE lAA-2 RADIATION SOURCE COMPARISON A I I % CORE INVENTORY RELEASED GROUP l R.G. 1.3 R.G. 1.7 Plant Design Basis AIR Noble Gases 100 100 100* Halogens 25 -- 25* All Remaining -- -- O WATER Noble Gases 0 -- Halogens -- 50 50** All Remaining -- 1 1** l l
- Uniformly mixed within the primary containment boundary
** Uniformly mixed in the suppression pool and reactor coolant O
1AA-30
% GESSAR II 238 NUCLEAR ISLAND TABLE lAA-3 CORE COOLING SYSTEMS AND AUXILIARIES 7
Division 1 Location Ecuipment Location D.G. Bldg. "A" Div 1 DG Engine Heat Exchanger
- DG Bldg. "B" Dig AB-1 Div 1 Elect. Switchgear AB-1 Dig E12-C002A RHR Pump "A" (LPCI) AB-4 'E12 AB-4 X73-ECUO4 RHR Pump "A" Room Cooler * (RIIR X73 I
(RIIR Room "A") Incl w/ pump RHR Pump " A" Seal Cooler, n om "B") In? E12-B001A RilR lleat Exchanger "A" (ElZ AB-2 [ E21-C001 LPCS PumI AB-2 'E13 (LPCS Room) X73-ECUO3 LPCS Pump Room Cooler * ( RilR (X73 Room "C") Ing FB-3 P53-AA001A Pneumatic Air Supply Receiver FB-3 P53 FB-3 P53-AA0012A Pneumatic Air Supply Air Bottles FB-3 PS3 FB-1 - Pneumatic Air Supply FB-1 - Non-essential Compressor & Dryer FPCCS Pump "A" I [' G41-C001A - lG41 FB-1 ( X63-ECUO2A FPCCS Pump Room Cooler "A"* FB-1 (X63 l
, G41-B001A FPCCS lieat Exchanger "A"* (G41 CT G41-F040A M.O. Valves + CT G41 CT G41-F044A M.O. Valver CT G41 FB-1 G41-N024A Pressure Transmitter FB-1 G41
- Equipment cooled by ESW System Div. 1.
** Equipment cooled by ESW System Div. 2.
*** Equipment cooled by IIPCS Service Water System
+ M.O. = Motor Operated l
1
)
r
22A7007/' Rev. .
- MAJOR EQUIPMENT .
Division 2 Division 3 L7uioment Location Equipment 2 DG Engina IIeat Exchanger ** IIPCS DG Bldg. HPC6 DG HX*** 2 Elect. Switchgear
-C002B R3R Pump "B" (LPCI)
ECUO7 RIR Pump "B" Room Cooler ** w/ pump RiiR Pump "B" Seal Cooler **
-B001B **
RdR lleat Exchanger "B" -C002C RIIR Pump "C" (LPCI) AB-2 ' E22-C001 'HPCS Pump ECUO6 RitR Pump "C" Room Cooler ** og ) ( X73-ECUO8 HPCS Pump Room L. w/ pump RIIR Pump "C" Room Cooler ** Cooler ***
-AA001B Pneumatic Air Supply Receiver -A A0 0 2B Pneumatic Air Supply Air Bottles Pneumatic Air Supply Non-essential Compressor & Dryer
-C001B FPCCU Pump "B" LECUO2B FP2CU Pump Room Cooler "B"**
-B001B ppCCU Ileat Exchanger "B"**
-F040B M.O. Valves I -F044B M.0. Valves
-N024B Pressure Transmitter i
1AA-31/lAA-32 #
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 TABLE lAA-4 CORE COOLING SYSTEMS AND AUXILIARIES VALVES AND INSTRUMENT TRANSMITTERS Equipment System MPL No. Description Location B21-Main Steam Valves A003 Air accumulators DW-1 (for SRV) A004 Air accumulators DW-1&2 (for SRV) F041 Safety releif valves DW-2 F047 Safety relief valves DW-2 F051 Safety relief valves DW-2 Pressure Transmitters N067 C,G,D,H HPCS CT-3 N073 A,B RHR A,B,C CT-3 N094 A,E,B,F RHR A,B,C; ADS A,B CT-3 N097 A,B RHR A,B,C ) CT-3 Transmitters N073 C,G,D,H RPV level & HPCS CT-3 N091 A,B,E,F RPV level & ADS A,B CT-3 RPV level & RHR/LPCI N095 A,B RPV level & ADS A,B CT-3 E12 RHR System Valves F008 Reactor to RHR CT F009 Reactor to RHR -DW F042 RHR-Hx to reactor CT-2 Flow Transmitters N013 SD cooling to RPV head AB*
'N015 RHR Discharge to reactor AB*
N052 RHR Discharge to reactor AB* Pressure Transmitters N053 RHR Pump discharge alarm AB* N055 RHR Pump discharge alarm AB* N056 RHR Pump discharge alarm AB* N057 RPV Recirc..to RHR pump AB* N058 A,B,C LPCI discharge to RPV AB* 1AA-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 _ TABLE lAA-4 CORE COOLING SYSTEMS AND AUXILIARIES l VALVES AND INSTRUMENT TRANSMITTERS (cont'd) Equipment System MPL No. Description Location E21 LPCS Valves F001 Suction valve AB F005 Discharge valve AB-4 Flow Transmitter N003 Pump discharge flow AB-2 Pressure Transmitter N053 Pump discharge pressure AB-2 E22 HPCS Valves F001 Cond. storage to pump ** AB-2 F004 Pump to RPV AB-2 F015 Supp. pool to pump ** AB-2 Pressure Transmitters N051 HPCS Pump discharge press. AB-2 N052 HPCS Pump suction press. AB-2 Flow Transmitter N005 HPCS Pump discharge flow AB-2 Level Transmitter M005 C,G Suppression pool level CT P33 Pneumatic Supply System DD005 Air filter CT-3 Valves Air Operated FF015 A,B Receiver to ADS FB FF017 A,B Receiver to ADS CT FF037 A,B Compressor to receiver FB FF038 A,B Air bottles to receiver FB PC Valve *** FF051 A,B Air bottles to receiver FB Press. Switch NN004 A,B Air bottle pressure switch FB
- RHR trcnsmitters are located in the respective RHR rooms
** Needed only to change source of pump suction
- PC = pressure control lAA-34
e ') Table 5 ( ) FISSION PRODUCT REMOVAL AND CONTROL System Equipment Location Division 1 Division 2 P38 SGTS Location Lccation MPL Code ~- MPL Code Exhaust Fan CC001A FB-5 CC001B FB-5 Heat Removal Fan CC003A FB-5 CC003B FB-5 SGTS Unit ZZ001A FB-4 ZZ001B FB-4
+
System Inlet Valve FF001A FB-4 FF001B ( AO) ( (++) ) FB-4 System Inlet Valve (MO) FF002A FB-4 FF002B FB-4 Air Intake (heat removal)(AO) FF003A FB-4 FF003B FB-4 Air Intake (heat removal)(MO) FF004A FB-4 FF004B FB-4 Exhaust Fan Suction (MO) FF006A FB-4 FF006B FB-4 N Heat Removal Fan Suction (MO) FF007A FB-4 FF007B FB-4 "
- 2:
Minimum Flow to Fan (MO) FF010A FB-4 FF010B FB-4 g. O Charcoal Water Drain (MO) FF015A FB-4 FF015B FB-4 $$ i Charcoal Water Spray (MO) FF016A FB-4 FF016B FB-4 $$ Decay Heat Removal Damper FF050A FB-5 FF050B FB-5 $[ Decay Heat Removal Damper FF051A FB-5 FF051B FB-5 Decay Heat Removal Damper FF052A FB-5 FF052B FB-5 U Decay Heat Removal Damper FF054A FB-5 FF054B FB-5 System Inlet Temperature NN603A FB-4 NN603B FB-4 System Differential Pressures RR012A FB-4 RR012B FB-4 System Differential Pressures RR013A FB-4 RR013B FB-4 System Differential Pressures NN019A FB-4 NN019B FB-4 Flow Through System, Flow NN02A FB-4 NN02B FB-4 Transmitter N X63 Fuel
,o Bldg. HVAC o
-a w SGTS room cooling unit
- EUCOlA FB-5 EUCOlB FB-5 t I
TABLE lAA-5 (Continued) FISSION PRODUCT REMOVAL AND CONTROL Location Division 1 Division 2 Location Location System Equipment MPL Code MDL Code E33 Main Steam Inboard System Inlet Valves (MO)++ F007 AB-7 Positive Leak- F008 AB-7 age Control Inboard System Pressure Control F002 AB Systems Inboard System Injection Valves F005 AB F007 AB (MO) F008 AB Outboard System Inlet Valves (MO) F027 AB-7 F028 AB-7 g H Outboard System Pressure Control F022 AB co $ Outboard System Injection Valves F025 AB h$ b (MO) F027 AB yy
- F028 AB NN Pressure Transmitter (Inboard) N001 AB-6 ss Pressure Transmitter (Inboard) N002 CT-3 yH Pressure Transmitter (Inboard) N003 CT-3 g Pressure Transmitter (Inboard) N004 AB-6 6 Pressure Transmitter (Inboard) N005 AB-6 Flow Transmitter (Inboard) N007 AB-6 Pressure Transmitter (Outboard) N021 AB-6 Pressure Transmitter (Outboard) N022 CT-3 Pressure Transmitter (Outboard) N023 CT-3 Pressure Transmitter (Outboard) N024 AB-6 Pressure Transmitter (Outboard) N025 AB-6 Flow Transmitter (Outboard) N027 AB-6 Control Panel H22-P074 AB-6 H21P073 AB-6 xU e5 0 Cooled by ESW *@
ww
+ Air operated ++ Motor operated 9 e e-
. . . _ . . _ . _ . _ . _ . _ . _ _ . _ _ _ _ _ . . _ . .. _ . . - _ . _ . _ _ - - . _ . - _ _ . _ _ _ _ . . _ . . - . _ _ . ~ . _ _ . _ - _ . . _ );
O O O $ TABLE lAA-5 (Continued) FISSION PRODUCT REMOVAL AND CONTROL i l Location i Division 1 Division 2 , Location Location System Equipment MPL Code MPL Code, ! .'P60 Water Positive _ Water Supply Tank AA-Oll FB-3 Dryer /Sep. Stg. Pool' Seal Isolation Water Supply . Valves (MO) FF034 FB FF002 FB ,
- ' Leakage Control FF035 AB FF003- AB -[
4 ' System -- FF036 AB FF004 AB
' FF049 AB FF005 AB FF056 AB FF020 AB ,
FF057 FB FF055 FB N 1 ESW--Tank-Fill FF026 FB-3 $ ! f' Condensate--Tank Fill FF027 FB-3 2 Pressurizing Air To Tank (PVC) PCV-031 FB-3 co 'E -Tank, Level Transmitter NN004 FB-3 OE
- y_ , Tank Pressure Transmitter NN003 FB-3 yy
+w w w.
" AB-9 P615 Air Positive. Compressor Package * - (non--essential) CC001A FB-3 CC001B ss AB-9 mH Air Receiver AA001A FB-3 AA001B Seal Leakage Isolation Control . ' Pressure Control (PVC) FF002 AB FF029 AB h-4 System- s Air -' Supply Valves . MO) ( FF004 AB FF010 AB- c.
r FF005 AB FF0ll AB FF006- AB FF058 AB P X73 Aux.' Bldg. HVAC- Elect.LArea A/C Unit ~ ACU02A AB-1 ACUO2B AB Self-Contained A/C Unit
- ACUO3 AB-1 ACUO4 AB-1 ,
l
*Co'oled b".ESW !
t
.. AO~ '= Air operated ~
~
i
!MO = Motor operated- w
- TT =~ Temperature transmitter y$
dPT =. Differential' pressure transmitter <~
'FT =-Flow transmitter *$
.w w r
L i R R 6
, n-.-,, .
TABLE lAA-6 COMBUSTIBLE GAS CONTROL Location Division 1 Division 2 Location Location System Equipment MPL Code MPL Code T41 Reactor Bldg. Shield Annulus Exh./ Rec. Fan CC004A FB-6 CC004B FB-6 HVAC Hydrogen Mixing Blower CC008A CT-4 CC008B CT-4 Drywell Bleed-Off Vent FF051 FB* -- -- System Manual valve Motor Operated Valve FF038 FB -- -- Manual Water Valves to H2 Mixing -- FB** 2 P41 Essential {FF125( CO Service Water Blower After Cooler and Oil Cooler IFF1281 Motor Operated Valves FF170 FB FFil4 AB O$ g $e T49 Flammability Thermal Recombiner Z001 CT-4 -- xx Power Supply Panel Z001 AB-6 -- ws $ Control System Z001 CB-1 -- Control Panel Hydrogen Recombiner ZZ001A CT-4 ZZ001B CT-4 g o Room Cooling Units for Shield ECUO3A FB-6 ECUO3B FB-6 X6 3 Fuel Building HVAC Building Exhaust Fan ***
- Manual valve is backup for MO-FF038. Note on P&ID says operation after accident should be by person wearing an air pack.
** Access to valves may also require an air pack.
*** Cooled by ESW. yU k -a e O 4
I I
# G e
utdsAR 11 22A7007 238 NUCLEAR ISLAND Rev. 7 , TABLE lAA-7 COMPARTMENT RADIATION DOSE RATE ( VS TIME AFTER ACCIDENT RADIATION DOSE RATE, R/HR VERSUS TIME AREA EVALUATED 4 hrs 1 day 5 days 30 days 100 days SECONDARY CONTAINMENT RHR Rooms 3 3 2 e AIRBORNE 5x10 3 6x10 2x10 4x10 5x10 4 4 e WATERBORNE 7x10 2x10 7x10 8x10 3x10 FUEL BUILDING e AIRBORNE 200 100 45 3 0.02 e WATERBORNE O O 0 0 0 0 CONTAINMENT SHINE + 0.5 0.02 0.008 0.002 < 0. 0 01 OUTSIDE SECONDARY CONT. ECCS Corridor
~- 0 AIRBORNE
- 50 60 8 2 0.4 e WATERBORNE
- 300 60 20 5 2 ELECT. EQUIP. ROOM e AIRBORNE
- 0 0 0 0 0 0 WATERBORNE
- 8 5 2 0.4 0.07 ASSUMPTIONS:
e R.G. 1.3 and 1.7 Source Terms 8% of secondary containment airborne leakage in each ECCS G compartment. e 5 GPM liquid leak for 5 days for ECCS compartments (l) e 10% of liquid leaked becomes airborne in ECCS rooms . e 1 AC/(2) day auxiliary building, 2/3 AC/(2) day fuel building , 8 No cleanup system in operation ( At 5 days, the flooding level is reached with consequent isolation. AC= air changes
- These sources are the sources within the ECCS rooms contributing to
() (~s radiation in the outside locations. There is no airborne contamination outside secondary containment.
+ Containment shine data also applies to areas outside secondary containment lAA-39
~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 . TABLE lAA-8 POST ACCIDENT RADIATION EXPOSURE, RADS Fluid Airborne Piping
- Leakage Containment Location Source Source Source Shine Total Auxiliary Building 6 6 6 IIPCS Room 1. 2x10 2. 8x10 8 . 6x10 22 5. 0x10 0 6 6 RHR A & B Room 1. 2x10 2. 6x10 8. 6x10 22 4 . 7x10 6 6 5 6 RHR Room C 1. 2x10 1. 6x10 8. 6x10 22 3. 7x10 6 6 5 6 LPCS Room 1. 2x10 2. 3x10 8. 6x10 22 4 . 4X10 4 3 4 ECCS Corridor ** 7x10 1. 4 x10 8. 8 x10 22 3.0x10 3 4 4 Above HPCS Room ** 2. 5 x10 9. 2x10 5.1x10 22 8. 6Xl0 EL (-) 6 ' -10 "
4 4 4 Above LPCS Room ** 2. 5 x10 1. 5 x10 5.1x10 22 9.1X10 EL (-) 6 ' -10 " Fuel Euilding 4 4 Operating Floor 1.8x10 22 1.8x10 4 4 FPCC Pump Room 1.8x10 22 1.8x10 4 4 Below Operating Floor 1.8x10 22 1.8x10 7 SGTS Room 1.4x10 22 1.4x10 _ Containment 7 CT 1 through 5 2.2x10
- Piping source is the average rather than the maximum for the compartment. Maximum value was used in exposure calculation.
- All radiation sources are within HPCS, RHR and LPCS rooms; ca3ealations were based on limiting room rather than making spec.fic calculations for each room.
lAA-40
. . _ _ _ . ~ ..
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 7 , (' TABLE lAA-9 COMPARISON OF CALCULATED EXPOSURES VS REQUIRED EXPOSURES FOR POST ACCIDENT SYSTEMS Required Post Accident Accident Location / Zone Equipment Exposure, Rads Exposure, Rads 3 4 Diesel Gen. Bldg. Diesel Generator 4.5x10 lx10 Auxiliary Building i HPCS Room HPCS Pump, Motor, 5.0 x 10 6 5 x 106 AB-2 Room Cooler, Valves,
; Inst., Actuators RHR A & B Room RHR Pump, Motor, Hx, 4.7 x 10 6 5 x 106 AB-4 Seal Cooler, Room i Cooler, Valves, Inst.,
Actuators RHR Room C RHR Pump, Motor, 3.7 x 106 5 x 106 AB-2 Room Cooler, Valves, Inst., Actuators
; LPCS Room LPCS Pump, Motor, 4.4 x 10 6 5 x 10 6 AB-2 Room Cooler, Valves,
Inst., Actuators -
ECCS Corridor 3.0 x 10 4 1.7'x 105 AB-6 Above HPCS Room El (-) 6 ' -10 " 8.6 x 10 4 -- Zone AB-1 Elect. Switchgear 1.7 x 10 5 Above LPCS Room i El (-) 6 ' -10" 9.1 x 10 4 --- Zone AB-1 Elect. Switchgear 1.7 x 10 5 l Fuel Building _ f Operating Floor 1.8 x 10" 2 x 10" \ I FPCC Pump Room 1.8 x 10" 2 x.10" i Below Oper. Floor 1.8 x 10" 2 x 10" SGTS Room 1.4 x 10 7 2 x 10 7 l I Control Room Control and Control Control-Instrumentation ( 1. x 10 <1 x 10 3 Equipment Rooms CB-1
- V Containment 7 7 CT-1 through CT-5 Instrument Transmitters 2.2 x 10 2.2 x 10 1AA-41/lAA-42 l
i
GESSAR II 22A7007 238 NUCLEAR ISII.ND Rev. 4 s e io O-10 _ I C 6 __ __ jf 10 _ DOSE RATE _ ' 8 104 ' :-- = 10 2 :
- ooss - -
- E h _
S E
~
2 2 y 102 _- _- 10 g
= :
5 O
=8 -
_ = - E, g 2 2 10 _-
-- 10 10' 19 _-
i 3o 0 i e i i l i til I i i i lit il i i i l i e til i iiiniill i i i 11 til goo 0 0.1 1.0 10 100 1000 TIME (days) Figure lAA-1. Diesel Generator Building 1 Radiation Versus Time Following LOCA 1AA-43
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 10 I O 105 g
/////////////////// QUALIFICATION EXPOSURE (rads) l -
l , ,.
.- - -A p s" 1
O 4 - s 3o 10 O INTEGRAT ED DOSE BASED ON
< AVERAGE PIPING DOSE RATE l
l 8 INTEGRATED DOSE BASED ON PEAK PIPING DOSE RATE I h INTEGRATED OSE BASED ON PEAK PIPING DG.'s RATE BUT , NO WATER LEAKAt : FROM b O \ OPERATING PUMP h 3 -\ - 10 5 E
- 3o \ O DOSE RATE BASED ON AVERAGE PlPING DOSE RATE g
2 5 8 bs 5
% s N E N
N N
% 4
% o ,, '
2 - 10 io i l l I 3 i 20 0 10 50 100 200 TIME (deys) Figure 1AA-2. ECCS Rooms - Dose Rate and Integrated Exposure, Post LOCA 1AA-44
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AB.2 GAS SAMPLES k Provision has been made to obtain gas samples from both the drywell and wetwell atmospheres and from the secondary containment atmosphere. The sample system is designed to operate over the range of potential pressures starting at one hour after a LOCA. Heat traced sample lines are used to prevent precipitation of moisture and resultant loss of iodine in the t ample lines. The gas samples may be passed through a particulate filter and silver zeolite cartridge for determination of particulate activity and total iodine activity by subsequent counting of the samples on a gamma spectrometer system. Alternately, the sample flow can bypass the iodine sampler and be chilled to remove moisture. A 15 milliliter grab _ sample can then be taken for determination of gaseous activity and for gas composition by gas chromatography. This size s/ sample has been adopted to be consistent with present o'ff-gas sample vial counting factors. Provision will be made in the laboratory to aliquot frac tions of the initial vial contents to other vials if the act.ivity is too high to count directly. O 1AB-3/ LAB-4 16E8
s
. 'GESSAR II .'
22A7007 238 NUCLEAR ISLAND . REV. 7 1AB.3 LIQUID SAMPLES (Contid) includes sample' coolers ~and control valves which select' liquid sample points. The station consists of a wall mounted
~
frame"and enclosures. _ Included within the sample station are equipment trays which contain modularized liquid and gas samplers. Each of these modules is approximately 18" x 14" x 20" high. The lower liquid sample portion of the sample station is shielded with 6 inches of lead brick, whereas the upper - gas sampler requires 2 incr.as of lead. The total weight of the wall mounted portion of the system is approximately 7000 pounds. The dimensions of the sample station including shielding is approximately 29" wide by 27" deep by 72" high. The frame is mounted so that the bottom of the frame is
)
i approximately 20 inches off the floor. The control 1 instrumentation is_ installed in a 2' by 4' by 6' high standard cabinet control panel. The panel contains the { conductivity, radiation level readouts, and the flow, , pressure, and temperature indicators,_and various control , valves and switches. The general front panel arrangement is shown in Figure _1AB.3-2. l
- Appropriate sample handling tools are included with the basic sample, station. A gas sampler vialN positioner and gas g vial cask $s-in'61uded. The gas vial is installed and removed..~
by use of the vial; positioner through the front of the gas sampler. Thekialisthe,nmanuallydroppedintothecask with the positioner swhich hilows the vial to be maintained about 3 feet ,from the individual performing the operation. The small volume liquid sample is remotely obtained through j the bottom of the sample : station by use of the small volume cask and cask positioner. The cask positioner holds the cask and positions the cask directly under the liquid sampler. v . x -
-s x LAB-7 -
(;. c % r - 16E9 s (s - l . l '"-
l GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 ) l 1AB.3 LIQUID SAMPLES (Cont'd) The sample vial is manually raised within the cask to engage the hypodermic needles. When the sample vial has been filled, the bottle is manually withdrawn into the cask. The sample vial is always contained within lead shielding during this operation. The cask is then lowered and sealed prior to transport to the laboratory. A large volume cask and cask positioner is used to remotely obtain the large vclume liquid sample. The positioner contains the cask and vial. The cask is transported to the required position under the sample station by a four wheel dolly cask positioner. When in position, this cask is hydraulically elevated approximately 1.5 inches by a small hand pump for contact with the sample station shielding under the liquid sample enclosure. The sample bottle is raised, held, and lowered by a simple push / pull cable. The cask is sealed by a threaded top plug inserted above the sample bottle. The weight of this large ] volume cask is approxiinately 700 pounds. The cask may be used for offsite shipment of the large volume sample; however, it will require additional packaging. A 15 milliliter bottle is contained within the lead shielded cask. This sample bottle is raised from its location in the cask to the sample station needles for bottle filling. The sample station will only deliver 10 milliliters to this sample bottle. When filled, the bottle is withdrawn into the cask. The sample bottle is always shielded by 5-6 inches of lead when in position under the sample station and during the fill and withdraw cycles, thus operator exposure is controlled. O 1AB-8 16E10
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AC.2 238 NUCLEAR ISLAND ASSESSMENT The 238 Nuclear Island design has been reviewed against the seven position items of NUREG-0737, Item II.E.4.2. The conclusions of that review follow: (1) SRP 6.2.4 Compliance Isolation provisions described in Table 6.2.4 were reviewed and found to meet the recommendations of NRC Standard Review Plan 6.2.4 (Rev. 1). (2) Essential vs. Non-Essential Classification The classification of essential and non-essential BWR systems has been defined in NEDO-24782. The results of that classification were used as the basis for the classification of systems in the 238 Nuclear Island O design, provided in Table 1AC.2-1. (3) Non-Essential System Isolation Non-essential systems are isolated by the containment isolation signals, and by redundant safety grade iso-lation valves. (4) Isolation Seal-In Logic Resetting isolation logic should not cause automatic reopening of containment isolation valves. Eight NSSS valves do not meet this criteria. Design changes to ensure compliance with the criteria are described in Subsection LAC.3 of this Attachment. O V 1AC-3 16I3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AC.2 238 NUCLEAR ISLAND ASSESSMENT (Cont'd) (5) Minimum Containment Pressure Setpoint The containment pressure setpoint has been reviewed by General Electric and the BWR Owners' Group and WP,3 found to be satisfactory. The results of that evaluation are described in a letter from D. B. Waters, Chairman BWR Owners' Group, to D. G. Eisenhut, (NRC), dated December 29, 1980 (Reference 20). (6) Sealed Isolation Design Isolation design provides for a sealed isolation function as discussed under Item (4), above. The normal operation purge lines meet the criteria of this position as l discussed in Subsection LAC.4 to this attachment. The 42" high purge supply and exhaust lines are isolated _ with a blind flange during normal operation. (7) Supply and Exhaust and Vent High Radiation Isolation 238 Nuclear Island supply and exhaust isolation valves are provided with a high radiation isolation signal. This signal is provided with an administrative-controlled manual override to mitigate a fuel handling accident or to provide access to the Mark III containment. O 1AC-4 16E11
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AC.3 MODIFICATIONS TO MEET POSITION ITEM 4 (} Position Item 4 states: "The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the auto-matic reopening of containment isolation valves. Reopening of containment isolation valves requires deliberate operator action". Clarification 4 to NUREG-0737, Item II.E.4.2, further states: Administrative provisions to close all isolation valves manually before resetting the isolation signals is not an acceptable method of meeting this position. The specific valves of the 238 Nuclear Island containment ] isolation system whose previously defined logic will allow automatic isolation valve reopening, if the operator acts to
~
[~ RESET the system's isolation valve control logic, are identified _ below. SYSTEM MPL # o RHR Sample Line E12-F060A/B E12-F075A/B-o Reactor Water Sample Lines B33-F019 B33-F020 o RCIC Steam Supply Lines E51-F063 E52-F064 The following subsections described the functional modifi-cations that are planned to the 238 Nuclear Island's present design to meet NUREG-0737, Item II.E.4.2 position Item 4 as previously discussed, LAC-5 16E12
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AC.3 MODIFICATIONS TO MEET POSITION ITEM 4 (Cont'd) 1AC.3.1 RHR and Reactor Water Sample Lines For these systems, the modification to the present design is to replace the existing two-position maintained contact switch for each valve with a three-position switch (momentary contact, spring return to " NORMAL" from both the "CLOSE" and "OPEN" mode positions) and to add two new relays for each valve circuit (Figure 1AC.3-1). These modifications in no way curtail the automatic primary isolation function initiation caused by the trip logic function. However, this modification does prevent the isolation valves from opening simultaneously as a consequence of resetting the trip logic function. Replacing the maintained contact switch with the new two-stage momentary contact switch assures the operator that, after initiating the trip logic reset function, the isolation valves will remain closed until deliberate control action is taken to individually open each valve. The added relays are utilized to assure separation of both the "OPEN" and "CLOSE" valve control mode functions. LAC.3.2 RCIC Steam Supply Lines For the RCIC system, the modifications to the present design adds another stage of contacts (7-8) to the existing single stage, two-position key lock maintained contact switch for each valve circuit (Figure LAC.3-2). In the present design, when the trip logic function is initiated by depressing the " RESET" push button, after all isolation signals have cleared, the isolation valves open simultaneously; there is no separation of the operator action to RESET from the operator action to begin opening the isolation valves. O 1AC-6 16I12
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1AC.3.2 RCIC STEAM SUPPLY LINER (Cont'd) This additional stage to the control switch prevents the isolation valves from opening simultaneously as a con-sequence of resetting the trip logic function. By adding another stage to the control switch with the same type of configuration (i.e., both stages currently close their contacts in the "OPEN" position switch mode), separation of the trip logic reset function from the deliberate operator action of opening the closed isolation valves is assured after an isolation has been cleared. Since the control switch must be in the "OPEN" position mode for normal RCIC operation, the added stage of closed contacts provided by the proposed design change effectively blocks the RESET function until the isolation valves are closed. To open the isolation valve (s), the operator must get the key for the control switch, reposition the switch to the O "CLOSE" mode and depress the " RESET" pushbutton; this action resets the isolation valve trip logic so that the trip logic function is armed and ready to respond to another system isolation, if required. But only when the operator returns the switch again to the "OPEN" position mode, can the i isolation valve ue opened under operator manual control. The design changes described above for the RHR Sample Line, Reactor Water Sample Lines and RCIC Steam Supply Lines provide the deliberate and separate operator actions required by NUREG-0737, Item II.E.4.2 position Item 4. O , 1AC-7/1AC-8 GEI-D1
. - _ - - _ _ -- . = _ _
i Tabl AC.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT , l r
.H On M System Essential Comments J
- 1. Reactor Head Cooling No Not a safety system.
- 2. Standby Liquid Control Yes- Should be available as back-up to CRD system.-
- 3. Low Pressure Coolant Yes Safety system.
N Injection w co O 4.. Separate 0;ppression Yes Main heat sink during isolation. am
*k rm
.E Pool Cooling My
* $:c ws
! mM '
- 5. Core Spray (High-Low Yes Safety systems.
Pressure)
- 6. Closed Cooling Water No Used for normal operation only. Not required for DBA but is necessary for the recirc, cleanup system operation, and fuel pool heat exchangers.
- 7. Containment Atmospheric Yes Combustible gas control function necessary Control to eliminate hydrogen / oxygen combustible atmosphere.
EgN 16J1 $$O a> O 4
Table LAC.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) o U A N System Essential Comments
- 8. Containment Spray Cooling Yes Necessary to control drywell/ containment pressure.
- 9. Automatic Depressurization Yes Safety system; control of RPV pressure.
w System g
$ bO o
Q 10. Standby Gas Treatment Yes Necessary to control emissions to g
$ environment. $
w H H
- 11. Auxiliary Building / Fuel Yes Necessary to cool safety system pumps and "
Building Emergency Cooling motors.
- 12. Reactor Core Isolation Yes Necessary for core cooldown following Cooling isolation from the turbine condenser and feedwater makeup.
- 13. Auxiliary Building / Fuel Yes/No If drain is required, the equipment is Building Equipment Drain probably out-of-service; check for indepen-dent isolation; drain should not back up and flood essential equipment.
Eww GEIC2, i
.# D o
4 O O O O
Table C.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) O 3 0 t i U System Essential Comments
- 14. Drywell and Containment No Not necessary for core cooldown.
, Floor Drains '
- 15. Emergency Service Water Yes Necessary to remove heat following System accident. Includes the ultimate heat to sink. w
$0 g 16. Instrument Air Yes Regarded as essential because this system pg supports safety equipment. Back-up $
accumulators are available for the safety [ equipment should the system fail. " k 8
- 17. Service Air No Serves no safety or shutdown function.
- 18. Main Steam Line No Not required for shutdown.
- 19. Feedwater Line No Not required for shutdown. Portion that I
is Class I is essential. GEIC3
$Do 4 0 4
Table 1AC.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) w H m System Essential Comments _
- 20. Reactor Water Sample No Not required for shutdown, but would be necessary for post-accident assessment.
Post-accident sample is a separate issue.
- 21. Control Rod Drive Cooling Yes No credit taken for reflood, but is desirable. U m
Not required during and immediately b nO
$i 22. Reactor Water Cleanup No F L1 following an accident. Necessary in MM w
long-term recovery. s r;m
- 23. Radwaste Collection No Not required for shutdown.
- 24. Recirculation System No Not required for jet pump plants because core can be cooled by natural circulation.
- 25. RHR Heat Exchangers Yes Main heat sink during isolation.
- 26. RHR Shutdown Cooling No Not essential but desirable to use if available. Not redundant, but safety grade.
Egu 16J4 %$O e4 O
~
O O O
O n Table m .C.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) , O M l 7 j $ System Essential Comments , i
- 27. RHR Vessel Head Spray No Not safety system.
1 I
- 28. RHR Containment Spray Yes Necessary to control pressure.
- 29. RHR - LPCI Function Yes Safety function. .
U
=
- 30. RHR - Steam Condensing No Not required as safety equipment. o Function n E
$ ta m M
w
" Ngw
- 31. Waste Collector and Surge No Not required for shutdown. w w m
Tank
- 32. Drywell Cooling No Used only in normal operation. Desirable to keep running.
- 33. Demineralized Water No Not assumed available in ECCS analysis.
- 34. Condensate Water No Not assumed available in ECCS analysis.
$U GEICS . . . .
$DO 4 O Q
Table 1AC.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) s G H System Essential Comments o
- 35. Fuel Pool Cooling No Boiling is acceptable, but make-up is necessary. Heat exchanges cooled by RBCCW system.
- 36. Drywell Bleed Yes Pressure control vent. Back-up to hydro-gen control.
w W m x 37. Positive Seal System Yes Insure that highly radioactive fluids are
> b ? confined in the reactor building. oO rm w Mm u
- 38. Traversing In-Core Probe (TIP) No Not required for reactor shutdown cooling. $$
m-
- 39. Fira Protection Yes Availability is essential, as the "acci-dent" may be the result of a fire.
- 40. Make-up Water Treatment No Serves no purposes during and immediately after accident. Longer-term availability necessary.
$l3 16J6 $>q
+8" G G G
GESSAR II 22A7007 238 NUCLEAR ISLAND RGv. 7 APPENDIX 1B CONTENTS Section Title Page IB ASSESSMENT OF UNRESOLVED SAFETY ISSUES 1B-1 1B.1
SUMMARY
1B-1 1B.l.1 Introduction 1B-1 ] 1B.1.2 Objective 1B-2 1B.1.3 238 Nuclear Island Applicability 1B-2 1B.2 UNRESOLVED SAFETY ISSUES 1B-3 1B.2.1 Introduction 1B-3 1B.2.2 Waterhammer (Task A-1) 1B-3 1B.2.2.1 Issue Description 1B-3 1B.2.2.2 NRC Activities 1B-4 O- 1B.2.2.3 Industry Activities and Resolution Status 1B-4 1B.2.3 Reactor Vessel Materials Toughness (Task A-ll) 1B-6 IB.2.3.1 Issue Description 1B-6 1B.2.3.2 NRC Activities 1B-7 1B.2.3.3 Industry Activities and Resolution Status 1B-7 1B.2.4 System Interaction in Nuclear Power Plants 1B-9 (Task A-17) 1B.2.4.1 Issue Description 1B-9 1B.2.4.2 NRC Activities 1B-9 1B.2.4.3 Industry Activities and Resolrtion Status 1B-11 1B.2.5 Safety Relief Valve Pool Dynamic Loads 1B-12 (Task A-39) 1B.2.5.1 Issue Description 1B-12 1B.2.5.2 NRC Activities 1B-14 1B.2.5.3 Industry Activities and Resolution Status 1B-14 i O 1B-i GEII-El
GESSAR II 22A7007 238 NUCLEAP ISLAND REV. 4 APPENDIX 1B CONTENTS (Continued) Section Title Page 1B.2.6 Seismic Design Criteria (Task A-40) 1B-14 1B.2.6.1 Issue Description 1B-14 1B.2.6.2 NRC Activities 1B-15 1B.2.6.3 Industry Activities and Resolution Status 1B-15 1B.2.7 Containment Emergency Sump Reliability 1B-16 (Task A-43) 1B.2.7.1 Issue Description 1B-16 1B.2.7.2 NRC Activities 1B-17 1B.2.7.3 Industry Activities and Resolution Status 1B-17 1B.2.8 Station Blackout (Task A-44) 1B-18 1B.2.8.1 Issue Description 1B-18 1B.2.8.2 NRC Activities 1B-19 1B.2.8.3 Industry Activities and Resolution Status 1B-20 1B.2.9 Snutdown Decay Heat Removal Requirements 1B-22 (Task A-45) 1B.2.9.1 Issue Description 1B-22 1B.2.9.2 NRC Activities 1B-23 1B.2.9.3 Industry Activities and Resolution Status 1B-23 1B.2.10 Safety Implications of Control Systems 1B-24 (Task A-47) IB.2.10.1 Issue Description 1B-24 1B.2.10.2 NRC Activities 1B-25 1B.2.10.3 Industry Activities and Resolution Status 1B-26 15-ii 118-A3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 APPENDIX 1B ASSESSMENT OF UNRESOLVED SAFETY ISSUES i 1B.1
SUMMARY
This appendix provides a summary of the relevant investiga-tive programs and measures utilized for addressing the unresolved safety issues applicable to the 238 Nuclear Island. Based on the information provided, it can be concluded that the 238 Nuclear Island can be operated without endangering the health and safety of the public. 1B.l.1 Introduction The NRC continuously evaluates the safety requirements used ( in its reviews against new information as it becomes avail-able. As new concerns or safety issues are identified, an assessment is conducted to determine the potential need for any immediate action which may be required to assure safe , operation. Depending on the results of the assessment, imme-diate licensing actions or changes in licensing criteria may or may not be necessary. In any event, further study by the NRC may be deemed appropriate to make judgments as to modifi-cation of NRC requirements or implementation of backfitting. I t These issues are sometimes referred to as " generic safety issues" because they are related to a particular class or l type of nuclear facility rather than a specific plant. The NRC has designated certain of these issues to be " unresolved
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safety issues" which they perceive as needing a more in-depth j technical review or study after the NRC staff has.made an , l initial determination that the safety significance does not prohibit continued operation or require immediate licensing ('N l
\- actions.
GEII-C 1B-1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.l.2 Objective The unresolved safety issues were initially identified in _ NUREG-0510 (Identification of Unresolved Safety Issues Relating to Nuclear Power Plants", January 1979). These _ issues are updated quarterly in NUREG 0606 (" Unresolved Safety Issues Summary"). The quarterly update provides current programmatic and schedule information and includes information relative to the implementation status of each issue for which technical resolution is complete. The overall objective of this appendix is to comply with the Atomic Safety and Licensing Appeal Board deci,sion (ALAB-444) that the Safety Evaluation Report (SER) for each plant should contain an assessment of each significant unresolved generic safety issue. The assessment should include a summary description of relevant investigative programs and the measures devised for dealing with the issues on the subject plant. 1B.l.3 238 Nuclear Island Applicability The unresolved safety issues outlined in NUREG 0606 include all issues for which technical resolution is not considered complete by the NRC. Several apply only to pressurized water reactors, one applies only to operating nuclear power plants, and one applies only to boiling water reactors with a Mark I Containment. The remaining unresolved safety issues which are applicable to the 238 Nuclear Island are given in Table 1B-1. The number of the generic task in the NRC program addressing each issue is given along with the section in which each issue is discussed. O GEII-C 1B-2
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.2.3 Industry Activities and Resolution Status * (Continued) In order to protect the 238 Nuclear Island emergency core cooling systems (Section 6.3.2.2.5) against the effects of waterhammer, the ECC systems are provided with jockey pumps. These jockey pumps keep the emergency core cooling system lines full of water up to the motor operated injection valves so that the emergency core cooling system pumps will not start pumping into voided lines. In addition, to ensure that the emergency core cooling system lines remain full, vents have been installed and filling procedures established. Further assurance for filled discharge piping is provided by pressure instrumentation that is used to initiate an alarm that sounds in the main control room if the pressure falls below a predetermined setpoint indicating difficulty main-taining a filled discharge line. Should this occur, or if an instrument becomes inoperable, the required action is identified in the Technical Specification. To provide additional protection against potential waterhammer events in the 238 Nuclear Island, piping design codes require [ consideration of impact loads. Approaches used at the design stage include: (1) avoiding rapid valve ~ operation; (2) piping layout to preclude water slugs in steam-filled lines; (3) use of snubbers and pipe hangers; and, (4) use of , vents and drains. The use of snubbers and pipe hangers are a by-product of protection from seismic loads,-however, their use helps _to mitigate the effects of waterhammer events. In addition, a preoperational vibration and dynamic effects test program will be conducted by the applicant, in conjunction with GE, in accordance-with Standard OM-3 of.the American Society of Mechanical Engineers.for all Class 1, Class.2, Class 3 and other piping systems and piping restraints. GEII-C 1B-5
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1B.2.2.3 Industry Activities and Resolution Status (Continued) These tests will provide adequate assurance that the piping restraints have been designed to withstand dynamic effects due to valve closures, pump trips, and other operating modes. Nonetheless, in the unlikely event that a pipe break did result from a severe waterhammer event, core cooling is assured by the emergency core cooling systems and protection is provided against the dynamic effects of such pipe breaks inside and outside of containment. In the event that the NRC's activities in Task A-1 identify any potentially significant waterhammer scenarios which have not explicitly been accounted for in the design and operation of the 238 Nuclear Island, corrective measures will be implemented. The task has not identified the need for measures beyond those already implemented. With respect to Task A-1, it is concluded that the 238 Nuclear Island can be operated without undue risk to the health and safety of the public. 1B.2.3 Reactor Vessel Materials Toughness (Task A-ll) 1B.2.3.1 Issue Description Because the possibility of failure of nuclear reactor pressure vessels (RPV) designed to the ASME Boiler and Pressure Vessel Code is remote, the design of nuclear facilities does not provide specific protection against reactor vessel failure. However, as plants accumulate more and more service time, neutron irradiation reduces the material fracture toughness and initial safety margins. Il8-C 1B-6
GESSAR II 22A7007 , 238 NUCLEAR ISLAND REV. 4 () 1B.2.3.1 Issue Description (Continued) Results from reactor vessel surveillance programs indicate that up to approximately 20 operating PWR's will have belt-line materials with marginal toughness, relative to the requirements of Appendices G and H of 10CFR Part 50, after comparatively short periods of operation. The NRC has concluded that for most plants now in the licensing process, current criteria, together with the materials currently employed, are adequate to ensure suitable safety margins for reactor vessels throughout their design lives. 1B.2.3.2 NRC Activities The principal objective of Task A-11 is to develop safety criteria to allow a more precise assessment of safety margins during normal operation, transients and accident conditions in those older reactor vessels that may have marginal fracture (} toughness. 1B.2.3.3 Industry Activities and Resolution Status Based upon evaluation of the 238 Nuclear Island reactor vessel materials toughness, it is concluded that adequate safety margins exist to assure brittle failure is avoided during operating, testing, maintenance, and anticipated transient conditions over the life of the unit. The 238 Nuclear Island complies with all requirements specified in 10CFR50 Appendicas G and H. A V 118-C -1B-7
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.3.3 Industry Activities and Resolution Status (Continued) The materials of the 238 Nuclear Island reactor vessel meet the fracture toughness requirements of NB-2300 of the ASME Code. Based on these requirements and the fabrication tech-niques employed by the vessel manufacturer, it is estimated that the total fluence over the design life would result in a final fracture toughness value above the minimum charpy impact requirement of 50 foot-pounds. This means that there is adequate toughness to avoid unstable crack growth from an existing defect at all times during design life. In addition, the surveillance program required by Appendix H of 10CFR Part 50 will afford an opportunity to reevaluate the fracture toughness periodically during a minimum of the first half of the design life. To assure adequate safety margins, adjustment to the nil ductility transition temperature (NDTT) and the development method for pressure / temperature curves are specified in 10CFR50 Appendices G and H. The amount of adjustment to the operating curves is a function of reference temperature, RT which depends upon the fast neutron (>l Mev) fluence NDT and copper and phosphorous content in the RPV material. For the 238 Nuclear Island, the copper and phosphorus content of the material is closely controlled. Furthermore, high upper shelf toughness is specified. The fast neutron fluence is low with respect to other reactor types because of the addi-tional moderator (water) in the annulus between the core shroud and the RPV. In addition, thermal shock followed by pressurization is not expected to occur in the 238 Nuclear Island. Furthermore, this conclusion was recently confirmed by the NRC staff in a NRC briefing on pressurized thermal shock on September 15, 1981. Therefore, the reactor pressure vessel material toughness (A-11) issue is not relevant to the 238 Nuclear Island. GEII-C IB-8
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1B.2.4.2 NRC Activities (Continued) laboratory contracts, is underway and a range of methods is being considered and tested for feasibility against a sample of some systems interaction candidates derived from Licensee Event Report evaluations. 1B.2.4.3 Industry Activities and Resolution Status The licensing requirements and procedures used in the 238 Nuclear Island safety reviews address many different types of systems interaction. Current licensing requirements are founded on the defense-in-depth principle. Adherence to this principle results in requirements such as physical separation and independence of redundant safety systems, and protection against events such as high energy line ruptures, missiles, high winds, flooding, seismic events, fires, operator errors, and sabotage. These design provisions supplemented by the current review procedures of the Standard Review Plan, which require inter-disciplinary reviews and which account, to a large extent, for review of potential systems interactions, provide for an adequately safe situation with respect to such interactions. The quality assurance program which is followed during the design, construction, and operational phases for each plant is expected to provide added assurance against the potential for adverse systems interactions. The development of systematic ways to identify and evaluate systems interactions may reduce the likelihood of common cause failures which could result in the loss of plant safety functions. However, operational experience with BWR plants that have many features similar to the 238 Nuclear Island indicates that the current review procedures and ll8-C 1B-ll
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.4.3 Industry Activities and Resolution Status (Continued) criteria described above, supplemented by post-TMI modifi-cations produce a 238 Nuclear Island design that is reason- . ably from the effects of potential systems interaction. In . addition, the Nuclear Safety Operational Analysis (Chapter 15) and the 238 Nuclear Island Probabilistic Risk Assessment, which considered common cause failures, confirm that such interactions are minor contributors to plant risk which has been shown to be significantly below that reported in WASH-1400. Applicants are expected to provide for a systematic visual inspection by a multidisciplinary team to review the "as-built" condition of the plant areas where physical interactions could potentially result in adverse effects on safety-grade equipment. Visual inspectionn of the plant are also expected g to be conducted by the applicaat to investigate spatially W coupled systems interactions that could be initiated by seismic events. Any spatial separations that do not meet established design criteria are to be reported for dis-position by analysis and/or hardware modification. With respect to Task A-17, it is concluded that the 238 Nuclear Island can be operated without endangering the health and safety of the public. 1B.2.5 Safety Relief Valve Hydrodynamic Loads (Task A-39) 1B.2.5.1 Issue Description All BWR/6 plants are equipped with a number of Safety / Relief Valves (S/RVs) to control primary system pressure transients. The S/RVs are mounted on the main steam lines inside the O GEII-C 1B-12
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.7.1 Issue Description (Continued) O The NRC concern addressed by this Task Action Plan as it - applies to boiling water reactors is primarily focused on - the potential for degraded ECCS performance as a result of thermal insulation debris that may be blown from pipes in , the drywell and by some means get into the suppression pool _ during a loss-of-coolant accident causing blockage of the pump suction lines. A second concern is potential vortex formation above the pump suctions and sWbsequent loss of net positive suction head to the ECCS pumps. 1B.2.7.2 NRC Activities The NRC is investigating the potential for debris from insulation causing blockage of the ECCS pump' strainers. The NRC investigation includes analysis of plant specific g designs and the types of insulation used. Also, the NRC had (_) conducted full scale containment emergency sump hydraulic tests at Alden Research Laboratory. The NRC's evaluation of the potential for void formation indicates that there is a much lower level of air-ingestion due to vortex formation than previously hypothesized by the NRC. The NRC has also found that up to 2 to 4 percent air void can be accommodated without significantly degrading pumping capacity. 1B.2.7.3 Industry Activities and Resolution Status With regard to potential blockage of the intake lines, it is very unlikely that insulation would be drawn into the ECCS pump suction lines. Insulation dislodged by a LOCA would primarily tend to collect in the region below the reactor inside the weirwall since this area is large in relation to the size of the annulus between the drywell and the weirwall. The debris in the drywell could only potentially by swept Oi into the suppression pool via the horizontal vents. GEII-C 1B-17
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1B.2.7.3 Industry Activities and Resolution Status (Continued) Insulation reaching the suppression pool would tend to either sink to the bottom or float on the surface of the pool. The ECCS suction strainers are sufficiently elevated above the bottom of the pool such that the fluid velocity in the direction of the suction strainers at the bottom of the pool is very low. This minimizes the potential for suction strainer plugging. Additionally, the ECCS suction strainers are twice as large as the size required to assure that adequate net positive suction head (NPSH) is available to the ECCS pumps. Thus, the ECCS suction strainer vould have to become more than 50 percent plugged before pump perfor-mance would be affected. The design is controlled such that the strainers will not become more than 50 percent plugged. The second concern, potential vortex formation, is not considered a serious concern for the Mark III containment due to the large depth of the pool and the low approach velocities. The 238 Nuclear Island has a minimum suction submergence for the ECCS systems of over 7 feet. With respect to Task A-43, it is concluded that the 238 Nuclear Island can be operated without endangering the health and safety of the public. 1B.2.8 Station Blackout (Task A-44) 1B.2.8.1 Issue Description Electrical power for safety systems at nuclear power plants must be supplied by at least two independent divisions. The systems used to remove decay heat and to cool the react.or core following a reactor shutdown are included among the g 118-C 1B-18
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.8.1 Issue Description (Continued) (} safety systems that must meet these requirements. Power sources for each electrical division for safety systems . include offsite alternating current power connections for normal supply and direct current battery charging, an onsite . standby emergency diesel generator for alternating current power supply and direct current battery cl.arging, and a stored energy direct current source (battery). The unlikely loss of all AC power (that is, loss of AC power from the offsite sources and from the onsite source) is ] referred to as station blackout. In the event of a staton blackout, the capability to cool the reactor core would be dependent on the timely restoration of AC power or the availability of those systems not requiring AC power. The NRC concern is over the probability and consequences of a station blackout event. [} 1B.2.8.2 NRC Activities Task A-44 involves a study of the following elements. First, the NRC through technical assistance contracts is evaluating the expected frequency and duration of offsite power loses at nuclear power plants. Next, an estimation of the reliability and an evaluation of the factors affecting ] the reliability of onsite emeryjency AC power supplies will be conducted. The risks to the public posed by station blackout events will then be evaluated. From the above information the NRC plans to assess the effectiveness of safety improvements they perceive may reduce public risk from station blackout events. O GEII-C 1B-19 _. . . - . _. ~ _ .-m ., , .e . . , , . - , , _ 9
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1B.2.8.2 NRC Activities (Continued) The issue of station blackout was considered by the Atomic Safety ad Licensing Appeal Board (ALAB-603) for the St. Lucie No. 2 facility. In addition, in view of the completion schedule for Task A-44 (October, 1982), the Appeal Board recommended that the Commission take expeditious action to accommodate a station blackout event. The commission has reviewed their recommendations and determined that some , interim measures should be taken at all facilities while Task A-44 is being conducted. NRC Generic Letter 81-04 requested a review and prompt implementation, as necessary, of emergency procedures and a training program for station blackout events. Consequently, interim emergency procedures and operator training for safe operation of the facility and restoration of alternating ',urrent power will be implemented
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by all operating reactors and by applicants prior to their fuel load date which will supplement the existing set of emergency procedure guidelines. 1B.2.8.3 Industry Activities and Resolution Status
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A loss of all offsite alternating current power involves a loss of both the preferred and backup sources of offsite power. The design basis, inspection and testing provisions for the offsite power system will be provided by the applicant in Section 8.2.
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The 238 Nuclear Island is provided with redundant power rapply systems to provide protection against the loss of _ offsite power. This includes three AC and four DC onsite power supply divisions. O GEII-C 1B-20
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 () 1B.2.9.2 NRC Activities The NRC objective is to develop a comprehensive and consis-tent set of shutdown cooling requirements, including the study of alternative means of shutdown decay heat removal and of diverse systems for this purpose. . The study will consist of a generic system evaluation and will result in recommendations regarding possible design requirements for improvements in existing systems. Also, an alternative decay heat removal method may be considared if it is evaluated to significantly reduce the overall risk to the public. r 1B.2.9.3 Industry Activities and Resolution Status -'
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The 238 Nuclear Island is designed with several alternative means for the removal of decay heat. The decay heat is l() normally rejected via the Power Conversion System. This l ! includes the supply of steam to the main turbine, heat being removed in the main condenser and condensate returned to the vessel by the feedwater system. If the condenser is not available, the safety relief valves operate in either an automatic or manual mode to discharge safety heat to the suppression pool with any of 13 pumps available to makeup . the subsequent loss in water inventory and the pool cooling system is operated to transfer this heat to the ultimate heat sink. Under normal shutdown conditions, the residual heat removal (RHR) system is effective in removing decay heat. During abnormal shutdown conditions, the water level in the RPV can ha reused to flood the steam lines and decay l heat can be removed via a safety / relief valve to the sup-pression pool and then transferred to the ultimate heat sink by use of the pool cooling system. These decay heat removal () and inventory makeup systems are summarized in Sections 15D.2 and are described in detail in Sections 5.4 and 6.3. GEII-C 1B-23 '..
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1B.2.9.3 Industry Activities and Resolution Status (Continued) Following the TMI accident, General diectric and the BWR Owners Group pe'rformed and documented extensive a'nalyses of feedwater transients and small-break loss-of-coolant accidents to support acceptability of current designs including the BWR/6. A report of these analyses was provided to the NRC in NEDO-24708A Revision 1, dated December, 1980. This report documents that adequate core cooling can be assured by the many diverne inventory maintenance and decay heat removal paths for a wide range of transients and accidents. The 238 Nuclear Island probabilistic risk assessment results
'N (Section 15D.3) indicate that the loss of long-term decay heat removal is not a dominant event. Consequently, improve-ments in the decay heat removal function would not significantly reduce the overall risk to the public. h With respect to Task A-45 it is concluded that the 238 Nuclear Island can be operated without endangering the health and safety of the public.
1B.2.10 Safety Implications of Control Systems (Task- A-47) 1B.2.10.1 Issue Description This issue concerns the potential for transients or accidents being m,ade more severe as a result of control system failures or malfunctions. These failures or malfunctions may occur independently or as a result of the accident or transient under consideration. . O n ll8-C 1B-24
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1B.2.10.3 Industry Activities and Resolution Status (} (Continued) A few carly operating boiling water reactors have experienced reactor vessel overfill transients with subsequent two-phase or liquid flow through the safety / relief valves. Following these early events, commercial-grade high-level trips (Level 8) have been installed in most BWRs including the 238 Nuclear Island to terminate flow from the appropriate systems. Periodic surveillance testing of these high level-trips is required by the Technical Specifications. No overfilling events have occurred since the Level 8 trips were installed. High level trips are also provided for the Reactor Core Isolation Cooling and High Pressure Core Spray systems. In addition, the 238 Nuclear Island has a high level scram that reduces the consequences of an overfill event. Both Nuclear Safety Operational Analyses (Chapter 15) and a PRA (Section, 15D.3% have been performed and they provide additional
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assurance that this issue is not a problem for the 238 Nuclear Island. With respect to Task A-47, it is concluded that the 238 Nuclear Island can be operated without endangering the health and safety of the public. 1B.2.11 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (Task A-48) 1B.2.11.1 Issue Description Postulated reactor accidents which result in a degraded or melted core may result in generation and release to the containment of large quantities of hydrogen. The hydrogen is formed from the reaction of the zirconium fuel cladding with steam at high temperatures and/or by radiolysis of f'] v ll8-C 1B-27
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1E.2.11.1 Issue Description water. Experience gained from the TMI-2 accident has , prompted the NRC to consider additional design provisions for handling larger hydrogen releases than those currently required by the regulations. 1B.2.11.2 NRC Activities In Task A-48 the NRC will investigate the means to predict the quantity and rate of hydrogen generation during degraded core accidents. In addition, the NRC will examine various means to cope with large releases of hydrogen to the con-tainment such as inerting the containment or controlled burning. The potential effects of proposed hydrogen control measures on safety, including the effects of hydrogen burns . on safety-related equipment, will also be investigated. Because of the potential for significant hydrogen generation as the result of an accident, 10CFR Section 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," and Criterion 41 of the General Design Criteria, " Containment Atmosphere Cleanup," in Appendix A to 10CFR Part 50, require that systems be provided to control hydrogen in the containment atmosphere following a postulated accident to ensure that containment integrity is maintained. The current regulation, 10CFR Section 50.44, requires that the combustible gas control system be capable of handling the hydrogen generated as a result of a design basis loss-of-coolant accident. To provide margin, the assumed hydrogen release is five times the amount calculated in demonstrating compliance with 10CFR Section 50.46 or the amount corresponding to reaction of the cladding to a depth of 0.00023 inch, whichever amount is greater. g OEII-C 1B-28
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 1
! 3B.7 SUPPRESSION POOL BASEMAT LOADS l
I In addition to the normal, seismic, deadweight and hydrostatic
, pressure loadings, that section of the basemat which forms the ; bottom of the suppression pool also experiences dynamic LOCA loads and oscillatory loads-during SRV actuation. The SRV loads are discussed in Attachment A.
j The outer half of suppression pool floor will experience a 10-psi
, bulk-pressure load associated with initial air-bubble formation as discussed in subsection 3B.6.1.3. This pressure rise above hydro-static is assumed to. increase to 21.8 psi at the drywell wall with the increase from 10 psi to 21.8 psi to be assumed'linenr and dis-1 tributed over 50% of the pool width as indicated in Figure 3B-67.
l This specification is based on the-observation that the maximum
- pressure that the initial bubble can ever have is the maximum dry-l well pressure during the accident. Data trace'no. 1 (Figure 3B-18) indicates that the pressure increase is no greater than 10 psi at
- a point halfway across the suppression pool. Thus, the specifica-tion that the pressure increases linearly between this point and
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i the drywell wall will bound the actual pressure distribution. .; , During the condensation and chugging phases of the postulated LOCA
- - blowdrwn, the loading on the basemat is the same as that on the
- containment'(Subsections 3B.6.1.9 and 3B.611.10). '
i
- The containment pressure increases to 3 psi due to drywell air
[ carryover and the long-term' pressure.and temperature increases l (Figure 3B-65). The time history of these pressure transients is shown on Figures 3B-55, 3B-66, and 3B-67. i SRV oscillating loads are defined in Attachment A.- The netiloading on the suppression pool. linear.will reverse during the negative pressure phase of-the oscillation and this lifting load on the liner needs to be considered during the design process. Where -
s ground water level is a concern,.this pressure is also a consider--
ation in the basemat. liner design.
i 3B-47/3B-48' _ _ , , - - _ ._ . -_._. - _ _ ._ - . . ~ . _ . __
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 _ SECTION 4.1 (J CONTENTS Section Title Page 4.1
SUMMARY
DESCRIPTION 4.1-1 4.1.1 Reactor Vessel 4.1-1 4.1.2 Reactor Internal Components 4.1-1 4.1.2.1 Reactor Core 4.1-2 4.1.2.1.1 Fuel Assembly Description 4.1-3 4.1.2.1.2 Assembly Support and Control Rod Location 4.1-3 4.1.2.2 Shroud 4.1-4 4.1.2.3 Shroud Head and Steam Separators 4.1-4 4.1.2.4 Steam Dryer Assembly 4.1-4 4.1.3 Reactivity Control Systems 4.1-4 4.1.3.1 Operation 4.1-4 4.1.3.2 Description of Control Rods 4.1-5 4.1.3.3 Supplementary Reactivity Control 4.1-5 4.1.4 Analysis Techniques 4.1-6
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4.1.4.1 Reactor Internal Components -4.1-6 4.1.4.1.1 TASA 4.1-6 4.1.4.1.2 DYSEA 4.1-7 4.1.4.1.3 HEATER 4.1-7 4.1.4.1.4 FAP-71 (Fatigue Analysis Program) 4.1-8 4.1.4.1.5 ANSYS 4.1-8 4.1.4.1.6 CLAPS 4.1-9 4.1.4.1.7 ASIST 4.1-10 4.1.4.2 Fuel Rod Thermal Analysis -4.1-10 , 4.1.4.3 Reactor Systems Dynamics 4.1-11 4.1.4.4 Nuclear Analysis 4.1-11 4.1.4.5 Neutron Fluence Calculations 4.1-11 4.1.4.6 Thermal-Hydraulic Calculations 4.1-12 . 4.1.5 References 4.1-13 , f'T J 4.1-i/4.1-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. "7 j O 4.1.3.1 Operation (Continued) counterbalance steam voids in the top of the core and effect ; significant power flattening. These groups of control elements, used for power flattening, experience a somewhat higher duty cycle and neutron exposure than the other rods in the control system. The reactivity control function requires that all rods be available for either reactor " scram" (prompt shutdown) or reactivity regula-tion. Because of this, the control elements are mechanically designed to withstand the dynamic forces resulting from a scram. They are connected to bottom-mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regu-lation or rapid scram insertion. The design of the rod-to-drive -' connection permits each blade to.be attached or detached from its drive without disturbing the remainder of the control system. The bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel open. 4.1.3.2 Description of Control Rods A description of the control rods is given in Subsection 4.2.2.4.1. 4.1.3.3 Supplementary Reactivity Control The core control requirements are met by use of the combined 3 effects of the movable control rods, supplementary burnable poison, and variation of reactor coolant flow. Description of- -- the supplementary burnable poison is provided in Sections 4.2 and 4.3. _ 4.1-5 o
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.1.4 Analysis Techniques 4.1.4.1 Reactor Internal Components Computer codes used for the analysis of the internal components are listed as follows: (1) TASA (2) DYSEA (3) HEATER (4) FAP-71 (5) ANSYS (6) CLAPS (7) ASIST - Detail description of these programs are given in the following sections. 4.1.4.1.1 TASA The TASA program is a two-dimensional and axisymmetric, transient, . nonlinear temperature analysis program. An unconditionally stable numerical integration scheme is combined with an iteration pro-cedure to compute temperature distribution within the body sub-jected to arbitrary time- and temperature-dependent boundary conditions. This program utilizes the finite element method. Included in the analysis are the three basic forms of heat transfer, conduction, radiation, and convection, as well as internal heat generation. In addition, cooling pipe boundary conditions are also treated. The output includes temperature of all the nodal points for the time instants specified by the user. The program can handle multitransient temperature input. O 4.1-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7
~N
)
4.1.4.1.1 TASA (Continued) A number of heat transfer problems related to the reactor pedestal have been satisfactorily solved using the program. 4.1.4.1.2 DYSEA
]
The DYSEA (Dynamic and Seismic Analysis) program is a GE proprie-tary program developed specifically for seismic and dynamic analy-sis of RPV and internals / building system. It calculates the dynamic response of linear structural systems by either temporal model superposition or response spectrum method. Fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass. Program DYSEA was based on program SAPIV with added capability to A ( ) handle the hydrodynamic mass effect. Structural stiffness and ,
%./
mass matrices are formulated similar to SAPIV. Solution is obtained in time domain by calculating the dynamic response mode-by mode. Time integration is performed by using Newmark's B-method. Response spectrum solution is also available as an option. It has been used extensively in all dynamic and seismic analysis of the RPV and internals / building system. i 4.1.4.1.3 HEATER HEATER is a computer program used in the hydraulic design of feed-water spargers and their associated delivery header and piping. The program utilizes test data obtained by GE using full-scale mockups of feedwater spargers combined with a series of models-which represent the complex mixing processes obtained in the
.. upper-plenum, downcomer, and lower plenum. Mass and-energy er ~
( balances throughout the nuclear steam supply system (NSSS) are N . modeled in detail. 4.'l-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.1.4.1.3 HEATER (Continued) - The program is used in the hydraulic design of the feedwater spargers for each BWR plant, in the evaluation of design modifi-cations, and the evaluation of unusual operational conditions. 4.1.4.1.4 FAP-71 (Fatigue Analysis Program) The FAP-71 computer code, or Fatigue Analysis Program, is a stress analysis tool used to aid in performing ASME-III Nuclear Vessel Code structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range and number of allowable fatigue cycles at points of interest, For structural locations at which the 3S, (P+Q) ASME Code limit is exceeded, the program can perform either (or both) of two elastic-plastic fatigue life evaluations: (1) the method reported in ASME Paper 68-PVP-3, or (2) the present method documented in Paragraph NB-3228.3 of the 1981 Edition of the ASME Section III Nuclear Vessel Code. The program can accommodate up to 25 transient strass states of as many as 20 structural Jocations. The program is used in conjunction with several shell analysis programs in determining the fatigue life of BWR mechanical com-ponents subject to thermal transients. 4.1.4.1.5 ANSYS ANSYS is a general-purpose finite element computer program designed to solve a variety of problems in engineering analysis. The ANSYS program features the following capabilities: (1) Structural analysis, including static elastic, plastic and creep, dynamic, seismic and dynamic plastic, and large deflection and stability analysis. 4.1-8
8 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.1.4.1.5 ANSYS (Continued) (2) One-dimensional fluid flow analysis. (3) Transient heat transfer analysis including conduction, convection, and radiation with direct input to thermal- ]
. stress analyses.
I (4) An extensive finite element library, including gaps, friction interfaces, springs, cables (tension only), direct interfaces (compression only), curved elbows, etc. Many of the elements contain complete plastic, creep, and swelling capabilities. (5) Plotting - Geometry plotting is available for all ele-ments in the ANSYS library, including isometric and perspective views of three-dimensional structures. 1 O (6) Restart Capability -- The ANSYS program has restart capability for several analyses types. An option is also available for saving the stiffness matrix once it is calculated for the structure, and using it for other loading conditions. ! ANSYS is used extensively in GE/NEBG for elastic and elastic-plastic analysis of the reactor pressure vessel, core support structures, reactor internals, fuel and fuel channel. l 4.1.4.1.6 CLAPS i CLAPS is a general-purpose, two-dimensional finite element program used to perform linear and nonlinear structural mechanics analysis. 4 The program solves plane stress, plane strain.and axisymmetric ! problems. It may be used to analyze for instantaneous pressure, s_/ temperature and flux changes,. rapid transients and steady-state, 4.1-9 L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.1.4.1.6 CLAPS (Continued) as well as conventional eleastic and inelastic buckling analyses of structural components subjected to mechanical loading. 4.1.4.1.7 ASIST The ASIST program is a General Electric code which can be used to obtain load distribution, deflections, critical frequencies and mode shapes in the "in-plane" or " normal-to-plane" modes for planar structures of any orientation that: (1) are statistically indeterminate; (2) can be represented by straight or curved beams; and (3) are under basically any loading, thermal gradient, or sinusoidal excitation. Deformations and resulting load distribu-tions are compared considering all strain energies (i.e., bending, torsion, shear and direct). ASIST also considers the effects of the deflected shape on loads and provides deflections calculated for the structure. In addition to this beam column (large deflec-tion) capability, the buckling instability of planar structures can also be calculated for the structure. In addition to this beam column (large deflection) capability, the buckling insta-bility of planar structures can also be calculated. The ASIST program has been used to determine spring constants, stresses, deflections, critical frequencies and associated modes shapes for frames, shafts, rotors, and other jet engine components. It has been used extensively as a design and analysis tool for various components of nuclear fuel assemblies. 4.1.4.2 Fuel Rod Thermal Analysis The fuel rod thermal analyses models are documented in Section 2 of Reference 2. 4.1-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O 4.1.4.3 Reactor Systems Dynamics The analysis techniques and computer codes used in reactor systems dynamics are described in Section 4 of Reference 1. Suu. :--
}
tion 4.4.4.6 also provides a complete stability analysis for the reactor coolant system. 4.1.4.4 Nuclear Analysis The analysis techniques are described and referenced in Section 3 of Reference 2. 4.1.4.5 Neutron Fluence Calculations Neutron vessel fluence calculations were carried out using a one-dimensional, discrete ordinates, Sn transport code with general anisotropic scattering. f~') v This code is a modification of a widely used discrete ordinates code which will solve a wide variety of radiation transport problems. The program will solve both fixed source and multi-plication problems. Slab, cylinder, and spherical geometry are allowed with various boundary conditions. The fluence calcula-tions incorporate, as an initial starting point, neutron fission distributions prepared from core physics data as a distributed source. . Anisotropic scattering was considered for all regions. The cross sections were prepared with 1/E flux weighted, P sub (L) matrices for anistropic scattering but did not include reson-ance self-shielding factors. Fast neutron fluxes at locations other than the core mid-plane were calculated using a two-dimensional, discrete ordinate code. The two-dimensional code is an extension of the one-dimensional code. O d 4.1-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.1.4.6 Thermal-Hydraulic Calculations Description of the thermal-hydraulic models are provided in Section 4 of Reference 2. 4.1.5 References
- 1. L. A. Carmichael and G. J. Scatena, " Stability and Dyanmic )
Performance of the General Electric Boiling Water Reactor," January 1977 (NEDO-21506).
- 2. " General Electric Standard Application for Reactor Fuel," ]
(NEDE-240ll-P-A, latest approved revision) . O l l O 4.1-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 SECTION 4.3 CONTENTS Section Title Page 4.3 NUCLEAR DESIGN 4.3-1 4.3.1 Design Bases 4.3-1 4.3.2 Description 4.3-1 4.3.2.1 Nuclear Design Description 4.3-1 4.3.2.2 Power Distribution 4.3-1 4.3.2.2.1 Power Distribution Calculations 4.3-1 4.3.2.2.2 Power Distribution Measurements 4.3-1 4.3.2.2.3 Power Distribution Accuracy 4.3-2 4.3.2.2.4 Power Distribution Anomalies 4.3-2 4.3.2.3 Reactivity Coefficients 4.3-2 4.3.2.4 Control Requirements 4.3-2 4.3.2.4.1 Shutdown Reactivity 4.3-3 s 4.3.2.4.2 Reactivity Variations 4.3-3
\_f 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3-3 4.3.2.6 Criticality of Reactor During Refueling 4 . 3 - -I 4.3.2.7 Stability 4.3-4 4.3.2.8 Vessel Irradiations 4.3-4 4.3.3 Analytical Methods 4.3-5 4.3.4 Changes 4.3-5 i 4.3.5 References 4.3-5 l
l i. I O l 4.3-i/4.3-li' I t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 l SECTION 4.3
)
TABLES i Table Title Page _ 4.3-1 Reference Core Loading Pattern Summary 4.3-6
~
4.3-2 Calculated Core Effective Multiplication and Control System Worth - No Voids, 20'C 4.3-7 4.3-3 Calculated Neutron Fluxes (Used to Evaluate Vessel Irradiation) 4.3-8 4.3-4 Calculated Neutron Flux at Core Equivalent Boundary 4.3-9 ILLUSTRATIONS Figure Title Page
~
4.3-1 Reference Core Loading Pattern 4.3-9a 4.3-2 Model for One-Dimensional Transport Analysis O~ of Vessel Fluence '4.3-10 v
-4.3-iil/4.3-iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O k/ 4.3 NUCLEAR DESIGN s See Appendix A, Section A.4.3 of Reference 1. 4.3.1 Design Bases See Appendix A, Subsection A.4.3.1 of Reference 1. 4.3.2 Description See Appendix A, Subsection A.4.3.2 of Reference 1. 4.3.2.1 Nuclear Design Description See Appendix A, Subsection A.4.3.2.1 of Reference 1. The ref-erence core loading pattern for the initial core is to be pro-vided by the applicant as shown in Figure 4.3-1. A summary of Os the fuel bundles loaded is shown in Table 4.3-1. 4.3.2.2 Power Distribution See Appendix A, Subsection A.4.3.2.2 of Reference 1. 4.3.2.2.1 Power Distri ' tion Calculations See Appendix A, subsection A.4.3.2.2.1 of Reference 1. A full range of calculated power distributions along with the resultant exposure shapes and the corresponding control rod pat-terns are shown in Appendix 4A for a typical BWR/6. 4.3.2.2.2 Power Distribution Measurements See Appendix A, Subsection A.4.3.2.2.2 of Reference 1. O-4.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 4.3.2.2.3 Power Distribution Accuracy See Appendix A, Subsection A.4.3.2.2.3 of Reference 1. I 4.3.2.2.4 Power Distribution Anomalies Stringent inspection procedures are utilized to ensure the correct rearrangement of the core following refueling. Although a mis-placement of a bundle in the core would be a very improbable event, calculations have been performed in order to determine the effects of such accidents on linear heat generation rate (LHGR) and criti-cal power ratio (CPR) . These results are presented in Chapter 15. The inherent design characteristics of the BWR are well suited to limit gross power tilting. The stabilizing nature of the large moderator void coefficient effectively reduces perturbations in the power distribution. In addition, the in-core instrumentation system, together with the on-line computer, provides the operator with prompt information on power distribution so that he can readily use control rods or other means to limit the undesirable effects of power tilting. Because of these design characteristics, it is not necessary to allocate a specific margin in the peaking factor to account for power tilt. If, for some reason, the power distribution could not be maintained within normal limits using control rods, then the operating power limits would have to be reduced as prescribed in Chapter 16 (Technical Specifications) . 4.3.2.3 Reactivity Coefficients See Appendix A, Subsection A.4.3.2.3 of Reference 1. 4.3.2.4 Control Requirements See Appendix A, Subsection A.4.3.2.4 of Reference 1. O 4.3-2
i l GESSAR II 22A7007 I 238 NUCLEAR ISLAND Rev. 7 l O 4.3.2.4.1 Shutdown Reactivity To assure that the safety design basis for shutdown is satisfied, an additional design margin is adopted: k-effective is calculated to be less than or equal to 0.99 with the control rod highest worth fully withdrawn. The cold shutdown margin for the reference core loading pattern is given in Table 4.3-2. 4.3.2.4.2 Reactivity Variations The excess reactivity designed into the core is controlled by the control rod system supplemented by gadolinia-urania fuel rods. , The gadolinia-urania concentrations for each fuel type are given , in Section 2 of Reference 1. _ O Control rods-are used during the cycle partly to compensate for burnup and partly to flatten the power distribution. 4 Reactivity balances are not used in describing BWR behavior be-cause of the strong interdependence of the individual constituents of reactivity. Therefore, the design process does not produce components of a reactivity balance at the conditions of interest. Instead, it gives the keff (Table 4.3-2) representing all effects combined. Further, any listing of components of a reactivity balance is quite ambiguous unless the sequence of the' changes is clearly defined. 4.3.2.5 Control Rod Patterns-and Reactivity Worths See Appendix A, Subsection A.4.3.2.5 of Reference 1. O 4.3-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4.3.2.6 Criticality of Reactor During Refueling See Appendix A, Subsection A.4.3.2.5 of Reference 1. 4.3.2.7 Stability See Appendix A, subsection A.4.3.2.6 of Reference 1. 4.3.2.8 Vessel Irradiations The neutron fluxes at the vessel have been calculated using the one-dimensional discrete ordinates transport code described in Subsection 4.1.4.5. The discrete ordinates code was used in a distributed source mode with cylindrical geometry. The geometry described six regions from the center of the core to a point beyond the vessel. The core region was modeled as a single homogenized cylindrical region. The coolant water region between the fuel channel and the shroud was described containing saturated water at 550 F and 1050 psi. The material compositions for the stainless steel in the shroud and the carbon steel in the vessel contain the mixtures by weight as specified in the ASME material specifications for ASME SA 240, 304L, and ASME SA 533 grade B. In the region between the shroud and the vessel, the presence of the jet pumps was ignored. A simple diagram showing the regions, dimensions, and weight fractions are shown in Figure 4.3-2. } The distributed source used for this analysis was obtained from the gross radial power description. The distributed source at any point in the core is the product of the power from the power description and the neutron yield from fission. By using the neu-tron energy spectrum, the distributed source is obtained for posi-tion and energy. The integral over position and energy i' normal-ized to the total number of neutrons in the core region. The core region is defined as a 1 centimeter thick disc with no transverse leakage. The power in this core region is set equal to the maximum power in the axial direction. 4.3-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 () 4.3.2.8 Vessel Irradiations (Continued) The neutron fluence is determined from the calculated flux by assuming that the plant is operated 90 percent of the time at 9 90 percent power level for 40 years or equivalent to 1 x 10 full power seconds. The calculated fluxes and fluence are shown in Table 4.3-3. The calculated neutron flux leaving the cylindrical core is shown in Table 4.3-4. 4.3.3 Analytical Methods See Appendix A, Subsection A.4.3.3 of Reference 1. 4.3.4 Changes See Appendix A, Subsection A.4.3.4 of Reference 1. () 4.3.5 References
- 1. " General Electric Standard Application for Reactor Fuel,"
(NEDE-240ll-P-A, latest approved revision). , l l l l l 4.3-5 l l I , _. . . . . . . , . , - . . , _ . . _ , . , , , . . , , , , . . - , _ . . - - , , , . . , _ _ . _ , . . , _ _ _ ., .. . - - . , - , _
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 I Table 4.3-1 REFERENCE CORE LOADING PATTERN l l Fuel Designation
- Number Loaded (Provided by Applicant) (Provided by Applicant)
%e O
4.3-6
GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rnv. 6 ' l 1 Table 4.3-4 ] l CALCULATED NEUTRON FLUX AT CORE EQUIVALENT BOUNDARY Lower Energy Flux Group Bound (eV) (n/cm2 -sec) 1 10.0 x 10' 3.6 x 10** 2 6.065 x 10' 5.3 x 10** 3 3.679 x 10' 2.0 x 10** 4 2.231 x 10' 3.9 x 10** 5 1.353 x 10 5 4.6 x 10** 6 8.208 x 10 5 4.1 x 10** 7 4.979 x 10 5 4.0 x 10** 8 3.020 x 10 8 2.8 x 10** 5 9 1.832 x 10 2.4 x 10** 10 6.738 x 10" 3'.4 x 10** , 11 2.479 x 10" 2.3 x 10** 12 9.119 x 10' 2.3 x 10** 13 3.355 x 10 8
.2.1 x.10**
14 1.234 x 10 8 2.1 x 10** 15 4.540 x 10* 2.0 x 10** 16 1.670 x 10* 2.1 x 10** . 17 6.144 x 10' 'l.9 x 10** ' l 18 2.260 x 10' l.9 x 10** 7 19' l.371 x 10' 9.2 x 10** 20 8.315 9.2 x 10** r 21 5.043 8.'4 x.10**- 22 3.059 8.7 x 10** 23 1.255 8.6-x 10'I 11 24 .l.125 8.5 x 10
~
l 25- 0.616 9.1'x 1011
-26 0.000 3.2 x 10 s.
1 l 4.3-9 l .- - -
GESSAR II 22A7007 238 NUCLr:AR ISLAND Rev. 7 0 (Provided by Applicant) O Figure 4.3-1. Reference Core Loading Pattern 4.3-9a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O 1 REACTOR CORE 4 WATER WA ER AR 3 SHROUD 5 VESSEL MATERIAL RADIUS VOLUME A 'ERAGE NO. NAME (inches) MATERIAL DENSITY 1 RE ACTOR COR E 92.58 WATER 0.318 g/cm3 UO2 2.334 g/cm3 O 2 304 L STAIN LESS STEE L ZlRCONIUM 0.056 g/cm3 C.978 g/cm3 WATER 99.9 WATER 0.74 g/cm3 3 SHROUD 101.9 304L STAINLESS STEE L FROM ASME SA 240 4 WATER 119.0 WATER 0.74 g/cm3 5 VESSEL 125.0 CARBON STEEL FROM ASME SA 533 6 AIR AIR 1.3 x 10-3g/cc Figure 4.3-2. Model for One-Dimensional Transport Analysis of Vessel Fluence _ 4.3-10
GESSAR II 22A7007 3 238 NUCLEAR ISLAND Rev, 7 Table 4.4-1 THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE 218-624 238-748 251-8C0 General Operating Conditions Reference rated thermal output (MWt) 2,894 3,579 3,833 Design power level for engineered safety features (MWt) 3,016 3,730 3 , 9 9 '- Rated steam flow rate, at 420 F final feedwater temperature (millions lb/hr) 12.453 15.400 16.49 Core coolant flow rate (millions lb/hr) 84.5 104.0 112.5 Feedwater flow rate (millions lb/hr) 12.428 15.367 16.46 System pressure, nominal in steam dome (psia) 1,040 1,040 1,040 j a i System pressure, nominal core y design (psia) 1,055 1,055 1,055 Coolant caturation temperature at core design pressure (*F) 551 551 551 Average hower density (kW/ liter) 52.4 54.1 54. 1 Maximum LHGR (kW/ft) 13.4 13.4 13.4 Average LHGR (kW/ft) 5.7 5.9 5.9 2 Core total heat transfer area (ft ) 61,151 73,303 78,398 Maximum he'at flux (Btu /hr-ft2) 361,600 361,600 361,600 Average heat flux (Btu /hr-ft ) 154,600 159,500 159,800 Design operating MCPR. See Table 15.0-2 i V 4.4-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 4.4-1 THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE (Continued)
~
218-624 238-748 251-800 General Operating Conditions (Continued) Core inlet enthalpy at 420 F FFWT (Btu /lb) 527.8 527.7 527.9 Core inlet temperature, at 420 F FFWT, ( F) 533 533 533 Core maximum exit voids within assemblies (%) 76.0 79.0 76.0 Core average void fraction, active coolant 0.411 0.414 0.412 Maximum fuel temperature ( F) 3,435 3,435 3,435 Active coolant flow area per assembly (in.2) 15.164 15.164 15.164 Core average inlet velocity (ft/sec) 6.82 6.98 7.07 Maximum inlet velocity (ft/sec) 7.90 8.54 8.57 Total core pressure drop (psi) 25.26 26.4 26.74 Core support plate pressure drop (psi) 20.84 22.0 22.32 Average orifice pressure drop Central region (psi) 5.41 5.71 5.78 Peripheral region (psi) 17.95 18.68 19.16 Maximum channel pressure loading (psi) 14.52 15.40 15.59 Average-power assembly channel pressure loading (bottom) (psi) 13.28 14.1 14.22 Shroud support ring and lower shroud pressure loading (psi) 24.84 25.7 25.12 Upper shroud pressure loading (psi) 4.0 3.7 2.8 O 4.4-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 i l i I l .I This page is intentionally left blank. 1 1 4 i e 4.4-13/4.4-14
4 , i-t ,
~
j j GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 7 -
4 Table 4.4-6 j a LENGTHS OF SAFETY INJECTION LINES 1 i i Nominal Diameter Pipe Length Loop Line (in) Schedule (ft) If 1 I I (Provided by Applicant.' i ) t 1 s : P a l [
.. k s.
6 D -
.O,.
, a s., . .
)% f
=1 4, s .
4.4-19/4.4-20 ,
t 2 4 GESSAR II [ 22A7007
' 238 NUCLEAR ISLAND Rev.'7 t
j APPENDIX 4A CONTENTS !
; Section Title Fage 1-
! 4A.1 INTRODUCTION ' ~4A.1-1 4 ! 4A.2 POWER DISTRIBUTION STRATEGY ~ - 4A.2-1 t r i i 4A.3 RESULTS OF CORE SIMULATION STUDIES , 4A.3-1 1
\
l 4A.4 REFERENCES - 4A.4-1 ] i 6 i l i i i i i t- ; I !
't
, b l. [- l 9 4 i
~ :- s
^
N 0,. _ t z.- ..,. . % -
=>
l
-O,:4, . 5 m , . 4A-i/_4A-if " m.' _
C 3:
~
_...,____m .- M~ i
_ = _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ . _ . _ _ . _ _ _ _ . __ _. . _ _ _ _ . . . _ . _ _ ._ __ j GESSAR II 22A7007 238 NUCLEAR ISLAND- Rev. 7 i () APPENDIX 4A ILLUSTRATIONS (Continued) i l Figure Title Page _ 4A-9b Relative Axial Power at 6.6 GWd/st j Cycle Exposure 4A.6-26 4A-9c Relative Axial Exposure at 6.6 GWd/st ! Cycle Exposure 4A.6-26 I 4A-9d Integrated Power per Bundle at 6.6 GWd/st Cycle Exposure 4A.6-27 l 4A-9e Average Bundle Exposure at 6.6 GWd/st i Cycle Exposure 4A.6-27 4A-10 Maximum Linear Heat Generation Rate as a Function of Cycle Exposure 4A.6-28 4A-ll Minimum Critical Power Ratio as a Function - of Cycle Exposure 4A.6-29 d 1 i O o e I i O 4A-V /4A-vi . l
-~ . . . . .. - - . - . . . . - - - . . . . , . - - - - - - _ _ - - . . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4A.2 POWER DISTRIBUTION STRATEGY O A basic operating principle used 'a minimize power peaking throughout an operating cycle has &een developed and is applied to boiling water reactors. The principle, the Haling principle, is described in Reference 1. The main concept is that "for any given set of end-of-cycle conditions, the power peaking factor is maintained at the minimum value when the power shape does not change.during the operating cycle". O . f p V ( 4A.2-1/4A'.2-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 4A.3 RESULTS OF CORE SIMULATION STUDIES O The following table itemizes the exposure step and its related figure numbers: Incremental Exposure (GWd/st) Sequence
- Figure Numbers 6.69 All-rods-out - Haling EOC 4A-la through 4A-1d I
0.2 A-2 4A-2a through 4A-2e 1.0 B-2 4A-3a through 4A-3e 2.0 A-1 4A-4a through 4A-4e 3.0 B-1 4A-Sa through 4A-Se 4.0 A-2 4A-6a through 4A-6e S.0 B-2 4 A-7a through 4 A-7e 6.0 A-1 4A-8a through 4A-Be 6.6 All rods out 4A-9a through 4A-9e s The detailed data presented demonstrates that this design can be
- ~ operated throughout this cycle with adequate margins to allow for operating flexibility. The variation of the maximum linear heat generation rate (MLHGR) with cycle exposure is presented in Figure 4A-10. Significant margin exists relative to the MLHGR -
operating limit. Maximum average planar linear heat generation , rates (MAPLHGR) are not calculated for this design since calcula-tions show the peak clad temperature (PCT) to be less than the 2200*F limit when the maximum single rod is at the 13.4 kW/ft limit. Adherence to the MLHGR limit will always assure meeting
~
the MAPLHGR limit. The variation of the minimum critical power rallo (MCPR) r/ith cycle exposure is shown in Figure 4A-ll. Sim- . ilarly, a large margin is indicated with respect to the expected l MCPR operating limit. . O)
\.
4A.3-1/4A.3-2 1
-m , , -
U} L.] ,) 1 1 2 3 4 5 6 1 8 9 10 11 12 13 14 15 J 1 0.3003 0.3751 0.4104 0.4224 0.4272 0.4216 2 0.3264 0.4470 0.6476 0.7892 0.8983 0.8501 0.9179 0.8340 3 0.3973 0.7105 0.8897 0.8405 1.0449 0.9271 1.0938 0.9306 0.9846 4 0.4135 0.7566 0.9627 0.9058 1.1126' O.9967 1.1841 1.0292 1.1840 0.9919 5 0.4228 0.7741 0.9921 0.9291 1.0382 1.0492 1.1179 1.0873 1.1582 1.0837 1.1220 6 0.4169 0.7790 1.0108 0.9478 1.1650 1.0415 1.2384 1.0224 1.2853 1.1098 1.2855 1.0328 7 0.4015 0.7645 1.0059 1.0239 1.0607 1.0728 1.1411 1.1122 1.1962 1.1506 1.2385 1.1460 1.1783 8 0.3299 0.7175 0.9717 0.9383 1.1657 0.9693 1.2345 1.0273 1.2973 1.0679 1.3422 1.1495 1.3314 1.0479 9 0.4507 0.8980 0.9138 1.0451 1.0396 1.1325 1.0255 1.1956 1.1518 1.2314 1.1653 1.2452 1.1481 1.1739 10 0.3038 0.6522 0.8489 1.1252 1.0482 1.2477 1.1135 1.2993 1.1351 1.3368 1.0769 1.3257 1.0722 1.3017 1.0298 11 0.3789 0.7970 1.0573 1.0108 1.1363 1.0354 1.2034 1.0739 1.2427 1.0854 1.2336 1.0652 1.2242 1.1085 1.1411 12 0.4152 0.9065 0.9373 1.1993 1.0999 1.3049 1.1610 1.3530 1.1803 1.3502 1.1549 1.3026 1.0385 1.2438 0.9801 $ , 13 0.4254 0.8553 1.1027 1.0377 1.1670 1.1156 1.2403 1.1534 1.2547 1.0836 1.2228 1.0462 1.1834 0.9901 1.0766 g3 14 0.4291 0.9217 0.9336 1.1891 1.0875 1.2876 1.1448 1.3304 1.1540 1.3135 1.1235 1.2648 1.0663 1.1812 0.9183 15 0.4221. 0.8340 0.9860 0.9923 1.1190 1.0310 1.1779 1.0487 1.1326 1.0378 1.1473 0.9920 1.0890 0.9220 0.9463 ;g C C1 (1 DG g, Mm p to u3 m Figure 4A-9d. Integrated Power per Bundle at 6.6 GWd/st Cycle Exposure $$ HH CO F4 z O I 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 J 1 25451.2 24479.5 24674.7 24015.8 24845.5 24938.7 2 25421.8 25008.9 3986.5 13835.1 5564.2 14365.2 5949.9 14615.4 3 24730.8 13488.2 5562.5 21248.2 6521.4 21941.6 7072.6 22318.8 15111.2 4
- 25454.8 13930.5 6198.4 20959.7 7230.6 21849.8 7636.5 22205.5 7879.8 22964.1 5 24960.7 14147.2 6456.7 22129.9 16057.2 20076.0 15368.1 21107.3 14923.6 21578.1 15781.5 6 25454.2 14110.3 6583.4 21864.7 7771.8 22141.2 8220.4 28390.7 8231.0 23279.3 0138.1 26609.4 7 24720.0 13907.1 6488.4 15959.0 16142.2 20022.9 15651.3 21540.0 14774.7 21506.6 14110.2 21493.9 15664.3 0- 25320.8 13411.0 4124.1 21760.8 7660.9 20696.2 8005.6 28590.3 8100.6 2s381.7 7802.9 23113.6 7646.6 27504.0 9 24979.5 5453.4 20828.6 15889.5 22290.0 15536.4 28565.3 14787.8 21388.5 14386.4 21633.2 14024.6 21715.0 16261.0 j 10 25284.8 3896.0 21117.1 7026.1 20653.9 7961.1 21486.9 8007.4 22964.5 7551.7 28395.6 7773.1 28031.4 7956.0 27843.9 6 11 24417.0 13724.6 6360.5 21505.3 14817.1 27894.4 14736.8 28240.4 14039.5 28136.9 13873.8 28543.2 12763.1 22294.5 16229.8 12 24443.1 5465.6 21756.3 7407.2 20950.7 7695.7 21294.4 7692.4 21323.0 7469.*, 21729.8 7865.1 28432.7 8212.6 28683.6 13 24781.0 14323.9 6972.3 22076,4 14833.1 23267.4 14308.8 23218.2 13994.0 20016.8 14038.6 28243.7 12814.2 28443.5 16382.4 14 24827.9 5904.2 22304.3 7835.10 21517.7 8182.1 21705.3 7910.7 21717.8 7847.4 21874.4 7979.9 22272.3 8074.4 28690.5 33 15 24946.7 14596.9 15061.0 2298D.1 16055.7 26829.3 15714.7 27506.3 15814.8 27365.3 16151.4 28011.3 16107.2 28452.3 21915.2 pg 33 m ),
<: -a
- C3 Average Bundle Exposure at 6.6 GWd/st Cycle Exposure o Figure 4A-9e. y
~
GESSAR II 22A7UOY' 238 NUCLEAR ISLAND Rev. 7 13.0 O 12.5 E n
- s IE
}a - .,
E W 12.0 - 5 A E 8
! O i
y 11.5 - a 3 E E
- s 11.0 -
1 l g o,3 I I I I I I I j_ 0 1 2 3 4 5 6 7 8 CYCLE EXPOSURE (GWdht) Figure 4A-10. Maximum Linear Heat Generation Rate as a Function of Cycle Exposure 4A.6-28 L . _ _ - - - - - - - -
_. _ . _ = . GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 1.50 0 1.45 - 2 6 3 9 4 e i. o - 6 8 _ b 5 P i2 o O 5 I iE 3 1.35 - l l l l 1.30 - l l I I I I I I I l 125 0 ,1 2 3 4 5 6 7 8 l CYCLE EXPOSURE (GWWud
. Figure 4A-11. Minimum Critical Power Ratio as a Function of Cycle Exposure 4A.6-29/4A.6-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev.'7
,7 I
j 6.3.2.8 Manual Actions (Continued) as indicating the operation of the ECCS. ECCS flow indication is the primary parameter available to assess proper operation of the system. Other indications, such as position of valves, status of circuit breakers, and essential power bus voltage, are available to assist him in determining system operating status. The electrical and instrumentation complement to the ECCS is discussed in detail in Section 7.3. Other available instrumentation is listed in the P& ids for the individual systems. Much of the monitoring instru-mentation available to the operator is discussed in more detail in Chapter 5 and Section 6.2. 6.3.3 ECCS Performance Evaluation The performance of the ECCS is determined through application of the 10CFR50 Appendix K evaluation models and then showing () conformance to the acceptance criteria of 10CFR50.46. Analytical models are documented in Subsection S.2.5.2 of Reference 4. The ECCS performance is evaluated for the entire spectrum of break sizes for postulated LOCAs. MAPLHGR results are for a fuel enrich-ment of approximately 3 wt% U-235. - The accidents, as listed in Chapter 15, for which ECCS operation is required are: Subsection Title 15.2.8 Feedwater Piping Break 15.6.4 Spectrum of BWR Steam System Piping Failures Outside of Containment 15.6.5 Loss-of-Coolant Accidents Chapter 15 provides the radiological consequences of the above listed events.
~
6.3-29
l GESSAR II 22A7007 238 NUCLEAR ISLAND R0v. 6 6.3.3.1 ECCS Bases for Technical Specifications The maximum average planar linear heat generation rates (MAPLHGR) calculated in this performance analysis provide the basis for Technical Specifications designed to ensure conformance with the acceptance criteria of 10CFR50.46. Minimum ECCS functional requirements are specified in Subsections 6.3.3.4 and 6.3.3.5, and testing requirements are discussed in Subsection 6.3.4. Limits on minimum suppression pool water level are discussed in Section 6.2. 6.3.3.2 Acceptance Criteria for ECCS Performance The applicable acceptance criteria, extracted from 10CFR50.46 are listed, and, for each criterion, applicable parts of Subsection 6.3.3 (where conformance is demonstrated) are indicated. A detailed description of the methods used to show compliance are shown in Subsection S.2.5.2 of Reference 4. _ Criterion 1: Peak Cladding Temperature "The calculated maximum fuel element cladding temperature shall . not exceed 2200 F." Conformance to Criterion 1 is shown in Subsections 6.3.3.7.3 (Break Spectrum), 6.3.3.7.4 (Design Basis Accident), 6.3.3.7.5 (Transition Break), 6.3.3.7.6 (Small Break), and specifically in Table 6.3-4 (MAPLHGR, maximum local oxidation, and peak cladding temperature versus exposure). Criterion 2: Maximum Cladding Oxidation "The calculated total local oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation." Conformation to Criterion 2 is shown in Figure 6.3-8 (break spectrum plot) , Table 6.3-4 (local oxidation versus cxposure) and Table 6.3-5 (break spectrum summary). 6.3-30
CESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 (D ( ,/ 6.3.3.2 Acceptance Criteria for ECCS Performance (Continued) Criterion 3: Maximum Hydrogen Generation "The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react." Conformance to Criterion 3 is shown in Table 6.3-5. Criterion 4: Coolable Geometry
" Calculated changes in core geometry shall be such that the core remains amenable to cooling." As described in Reference 2, Section III .A, conformance to Criterion 4 is demonstrated by conformance to Criterion 1 and 2.
73 ) t N_/ Criterion 5: Long-Term Cooling "After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an accept-ably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core." Conformance to Criterion 5 is demonstrated generically for General Electric BWRs in Reference 2, Section III.A. Briefly summarized, the core remains covered to at least the jet pump suction elevation and the uncovered region is cooled by spray cooling and/or by steam generated in the covered part of the core. 6.3.3.3 Single-Failure Considerations 73 The functional consequences of potential operator errors and
- i
\_ / single failures (including those which might cause any manually 6.3-31
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.3.3.3 Single-Failure Considerations (Continued) controlled electrically operated valve in the ECCS to move to a position which could adversely affect the ECCS) and the potential for submergence of valve motors in the ECCS are discussed in Subsection 6.3.2. There it was shown that all potential single failures are no more severe than one of the single failures identified in Table 6.3-3. It is therefore only necessary to consider each of these single failures in the ECCS performance analyses. For large breaks, failure of one of the diesel generators is, in general, the most severe failure. For small breaks, the HPCS is the most severe failure. A single failure in the ADS (one ADS valve) has no effect in large breaks. Therefore, as a matter of calculational convenience, it is assumed in all calculations that one ADS valve fails to operate in addition to the identified single failure. This assumption reduces the number of calculations required in the performance analysis and bounds the effects of one ADS valve failure and HPCS failure by themselves. The only effect of the assumed ADS valve failure by the calculations is a small increase (on the order of 100 F) in the calculated temperatures following small breaks. 6.3.3.4 System Performance During the Accident In general, the system response to an accident can be described as: (1) receiving an initiation signal; (2) a small lag time (to open all valves and have the pumps up to rated speed); and (3) finally, the ECCS flow entering the vessel. 6.3-32
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6
/~'i 6.3.3.4 System Performance During the Accident (Continued)
U Key ECCS actuation setpoints and time delays for all the ECCS systems are provided in Table 6.3-1. The minimization of the delay from the receipt of signal until the ECCS pumps have reached i rated speed is limited by the physical constraints on accelerating the diesel-generators and pumps. The delay time due to valve motion in the case of the high pressure system provides a suitably conservative allowance for valves available for this application. In the case of the low pressure system, the time delay for valve motion is such that the pumps are at rated speed prior to the time the vessel pressure reaches the pump shutoff pressure. The flow delivery rates analyzed in Subsection 6.3.3 can be deter-3 mined from the head-flow curves in Figures 6.3-3, 6.3-4 and 6.3-5 and the pressure versus time plots discussed in Subsection 6.3.3.7. Simplified piping and instrumentation and process diagrams for () the ECCS are referenced in Subsection 6.3.2. The operational sequence of ECCS for the DBA is shown in Table 6.3-2. Operator action is not required, except as a monitoring function, during the short-term cooling period following the LOCA. During the long-term cooling period, the operator will take action as specified in Subsection 6.2.2.2 to place the containment cooling system into operation. 6.3.3.5 Use of Dual Function Components for ECCS See Appendix A, Subsection A.6.3.3.5 of Reference 4. O 6.3-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 7 O 6.3.3.6 Limits on ECCS System Parameters See Appendix A, Subsection A.6.3.3.6 of Reference 4. O 6.3.3.7 ECCS Analyses for LOCA 6.3.3.7.1 LOCA Analysis Procedures and Input Variables See Appendix A, Subsection A.6.3.3.7.1 of Reference 4. The sig-nificant input variables used by the LOCA codes are listed in Table 6.3-1 and Figure 6.3-9. _ O 6.3-34
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 6.3.3.7.5 Transition Recirculation Line Break Calculations (Continued) (11) fuel rod convective heat transfer coefficients (small break methods) as a function of time; and (12) peaking cladding temperature (small break methods) as a function of time. 6.3.3.7.6 Small Recirculation Line Break Calculations Important variables from the analysis of the small break yielding the highest cladding temperature are shown in Figures 6.3-48 through 6.3-51. These variables are: 4 (1) water level as a function-of time; (2) pressure as a function of time; (3) convective heat transfer coefficients as a function of time; and (4) peak cladding temperature as a function of time. i The same variables resulting from the analysis of a less limiting small break are shown in Figures 6.4-52 through 6.3-55. l 6.3.3.7.7 Calculations for other Break Locations l Reactor water level and vessel pressure and peak cladding temperature and fuel rod convective heat transfer coefficients are shown in Figures 6.3-56 through 6.3-59 for the core spray- _ line break, Figures 6.3-60 through 6.3-63 for the feedwater l p) (, line break, and in Figures'6.3-64 and 6.3-65 for the main steamline break inside the containment. 1 ( l 6.3-41 n - ,--
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 6.3.3.7.7 Calculations for Other Break Locations (Continued) An analysis was also done for the main steamline break outside the containment. Reactor water level and vessel pressure and peak cladding temperature and fuel rod convective heat transfer coefficients are shown in Figures 6.3-68 through 6.3-71. 6.3.3.7.8 Improved Decay Heat Correlation Section I.A.4 of 10CFR50, Appendix K, requires use of the 1971 ANS Standards Subcommittee proposed decay heat standard for ECCS licensing evaluations. The current method for applying the 1971 standards in BWR LOCA calculations is outlined in GE's approved ECCS evaluation model (Reference 2) . In 1979, the American ) National Standards Institute approved and the ANS published a much improved decay heat standard (Reference 3). A detailed ] technical basis for an improved GE BWR decay heat correlation based on the 1979 standard is outlined in Appendix 6A. Use of h the improved correlation in the currently approved GE LOCA models will provide increased ECCS criteria margins. , Application of the correlation described in Appendix 6A is optional. To use it in place of the current method, a utility must provide the NRC with a request for exemption from Section I.A.4 of 10CFR50, Appendix K. The utility must reference Appendix 6A as the technical justification for the exemption. 6.3.3.8 LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is concluded that the ECCS will perform its function in an acceptable manner and r.eet all of the 10CFR50.46 acceptance criteria, given operation at or below the MAPLHGRs in Table 6.3-4. O l 6.3-42
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 1 ()6.3.4.2.4 LPCI Testing Each LPCI loop can be tested during reactor operation. The test conditions are tabulated in Figures 6.3-4a, b and c. During plant operation, this test does not inject cold water into the reactor because the injection line check valve is held closed by vessel pressure, which is higher than the pump pressure. The injection line portion is tested with reactor water when the reactor is shut down and when a closed system loop is created. This prevents unnecessary thermal stresses. To test an LPCI pump at rated flow, the test line valve to the suppression pool is opened, the pump suction valve from the suppression pool is opened (this valve is normally open) and the pumps are started using the remote / manual switches in the control room. Correct operation is determined by observing the instruments in the control room. If an initiation signal occurs during the test, the LPCI System returns to the operating mode. The valves in the test bypass lines are closed automatically to assure that the LPCI pump discharge is correctly routed to the vessel. 6.3.5 Instrumentation Requirements Design details including redundancy and logic of the ECCS instrumentation are discussed in Section 7.3. All instrumentation required for automatic and manual initiation of the HPCS, LPCS, LPCI and ADS is discussed in Subsection 7.3.1 and is designed to meet the requirements of IEEE-279 and other applicable regulatory requirements. The HPCS, LPCS, LPCI and ADS can be manually initiated from the control room. 6.3-47
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 6.3.5 Instrumentation Requirements (Continued) The HPCS, LPCS and LPCI are automatically initiated on low reactor water level or high drywell pressure. (See Table 6.3-8 for speci-fic initiation levels for each system) The ADS is automatically actuated by sensed variables for reactor vessel low water level and drywell high pressure plus indication that at least one LPCI or LPCS pump is operating. The FPCS, LPCS and LPCI automatically return from system flow test modes to the emergency core cooling mode of operation following receipt of an automatic initiation signal. The LPCS and LPCI system injection into the RPV begin when reactor pressure decreases to system discharge shutoff pressure. HPCS injection begins as soon as the HPCS pump is up to speed and the injection valve is open, since the HPCS is capable of injecting water into the RPV over a pressure range from 1177 psid* to 200 3 psid , 6.3.6 References
- 1. H.M. Hirsch, " Methods for Calculating Safe Test Intervals and Allowable Repair Times for Engineered Safeguard Systems", January 1973 (NEGO-10739) . _.
- 2. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix", -
November 1975 (NEDE-20566P).
- 3. " Decay Heat Power in Light Water Reactors", ANSI /ANS 5.1-1979, Approved by American National Standards Instituce, August 29, 1979.
- 4. " General Electric Standard Application for Reactor Pael-United States Supplement", NEDE-20411-P-A-US (latest approved revision). ,
3 psid - differential pressure between RPV and pump suction source. 6.3-48
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 SECTION 15.1 () TABLES Table Title Page 15.1-1 Sequence of Events for Figure 15.1-1 15.1-17 15.1-2 Sequence of Events for Figure 15.1-2 15.1-18 15.1-3 Sequence of Events for Figure 15.1-3 15.1-19 15.1-4 Sequence of Events for Figure 15.1-4 15.1-20 15.1-5 Sequence of Events for Inadvertent Safety / Relief Valve Opening 15.1-21 15.1-6 Sequence of Events for Inadvertent RHR Shutdown Cooling Operation 15.1-22 . ILLUSTRATIONS Figure Title Page. 15.1-1 Loss of 100'F Feedwater Heating- 15.1-23 15.1-2 Loss of 100'F Feedwater Heating 15.1-24 15.1-3 Feedwater Controller Failure Maximum Demand, with High Water Level Trips 15.1-25 15.1-4 Pressure Regulatory Failure Open 130% 15.1-26 , l O 15.1-v/15.1-vi
GESSAR II 22A7007 238 NUCLFAR ISLAND Rov. 6
/N 15.1.4.2.2 Systems Operation U
This event assumes normal functioning of normal plant instrumentar tion and controls, specifically the operation of the pressure regulator and level control systems. 15.1.4.3 Core and System Performance _. The opening of a S/R valve allows steam to be discharged into the suppression pool. The sudden increase in the rate of steam flow leaving the reactor vessel causes a mild depressurization transient. The pressure regulator senses the nuclear system pressure decrease 4 and within a few seconds closes the turbine control valve far enough to stabilize reactor vessel pressure at a slightly lower value and reactor power settles at nearly the initial power level. () Thermal margins decrease only slightly through the transient, and no fuel damage results from the transient. MCPR is essentially unchanged and, therefore, the safety limit margin is unaffected ~ and this event does not have to be reanalyzed for specific core configurations. 15.1.4.4 Barrier Performance i As discussed above, the transient resulting from a stuck open relief valve is a mild depressurization which is within the range of normal load following and therefore has no significant effect - on RCPB and containment design pressure limits. 15.1.4.5 Radiological Consequences While the consequences of this event does not result in fuel fail-ure, it does result in the discharge of normal coolant activity to the suppression pool via SRV operation. Since this activity is b(~'N 15.1-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 15.1.4.5 Radiological Consequences (Continued) contained in the primary containment, there will be no exposures to operating personnel. Since this event does not result in an uncontrolled release to the environment, the plant operator can chocse to leave the activity bottled up in the containment or discharge it to the environment under controlled release con-ditions. If purging of the containment is chosen, the release will be in accordance with the established technical specifica-tions; therefore, this event, at the worst, would only result in a small increase in the yearly integrated exposure level. 15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR This event is not applicable to BWR plants. 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.6.1 Identification of Causes and Frequency Classification 15.1.6.1.1 Identification of Causes At design power conditions, no conceivable malfunction in the shut-down cooling system could cause temperature reduction. In startup or cooldown operation, if the reactor were critical or near critical, a very sicw increase in reactor power could result. A shutdown cooling malfunction leading to a moderator temperature decrease could result from misoperation of the cooling water con-trols for the RHR heat exchangers. The resulting temperature decrease would cause a slow insertion of positive reactivity into the core. If the operator did not act to control the power level, a high neutron flux reactor scram trould terminate the transient without violating fuel thermal limits and without any measurable increase in nuclear system pressure. 15.1-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 7 () 15.1.6.1.2 Frequency Classification Although no single failure could cause this event, it is conserva-tively categorized as an event of moderate frequency. 15.1.6.2 Sequence of Events and Systems Operation 15.1.6.2.1 Sequence of Events A shutdown cooling malfunction leading to a moderator temperature decrease could result from misoperation of the cooling water con-trols for RHR heat exchangers. The resulting temperature decrease causes a slow insertion of positive reactivity into the core. Scram will occur before any thermal limits are reached if the oper-ator does not take action. The sequence of events for this event is shown in Table 15.1-6.
) 15.1.6.2.2 System Operation A shutdown cooling malfunction causing a moderator temperature decrease must be considered in all operating states. However, this event is not considered while at power operation since the nuclear system pressure is too high to permit operation of the shutdown cooling (RHRs).
No unique safety actions are required to avoid unacceptable safety results for transients as a result of a reactor coolant temperature decrease induced by misoperation of the shutdown cooling heat exchangers. In startup or cooldown operation, where the reactor is at or near critical, the slow power increase resulting from the cooler moderator temperature would be controlled by the operator in the same manner normally used to control power in the source or intermediate power ranges. O (_) 15.1-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 15.1.6.3 Core and System Performance h The increased subcooling caused by misoperation of the RHR shut-down cooling mode could result in a slow power increase due to the reactivity insertion. This power rise would be terminated by a flux scram before fuel thermal limits are approached. Therefore, -. only qualitative description is provided here and this event does not have to be analyzed for specific core configurations. 15.1.6.4 Barrier Performance As noted above, the consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel or containment are designed; therefore, these barriers maintain their integrity and function as designed. 15.1.6.5 Radiological Consequences Since this event does not result in any fuel failures, no analysis of radiological consequences is required for this event. O 15.1-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 i Table 15.1-1 SEQUENCE OF EVENTS FOR FIGURE 15.1-1 Time (sec) Event l 0 Initiate a 100*F temperature reduction in the feedwater system. 5 Initial effect of unheated feedwater starts to raise core power level but AFC system automat-ically reduces core flow to maintain initial steam flow. 100 Reactor variables settle into new steady state. O O 15.1-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.1-2 SEQUENCE OF EVENTS FOR FIGURE 15.1-2 Time (sec) Event 0 Initiate a 100 F temperature reduction into the feedwater system. 5 Initial effect of unheated feedwater starts to raise core power level and steam flow. 7 Turbine control valves start to open to regulate pressure. 36 APRM initiates reactor scram on high thermal power. 44.0 Narrow Range (NR) sensed water level reaches Level 3 (L3) setpoint. Recirculation pumps tripped to low frequency speed. >50 (est) Recirculation Pump Trip initiated due to Level 2 Trip. (not included in simulation). >50 (est) Wide Range (WR) sensed water level reaches Level 2 (L2) setpoint. >80 (est) HPCS/RCIC flow enters vessel (not simulated). >90 (est) Reactor variables settle into limit cycle. O 15.1-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 () Table 15.1-3 SEQUENCE OF EVENTS FOR FIGURE 15.1-3 i Time (sec) Event 0 Initiate simulated failure of 130% upper limit at - system design pressure of 1065 psig on feedwater flow.
)
11.8 L8 vessel level setpoint initiates reactor scram and trips main turbine and feedwater pumps, 11.9 Recirculation pump trip (RPT) actuated by stop valve position switches. l 11.9 Main turbine bypass valves opened due to turbine trip. 13.2 Safety / relief valves open due to high pressure. 18.2 Safety / relief valves close.
>20 (est) Water level dropped to low water level setpoing (Level 2).
,- O >50 (est) RCIC and HPCS flow into vessel (not simulated). s t O 15.1-19
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 7 Table 15.1-4 h SEQUENCE OF EVENTS FOR FIGURE 15.1-4 Time (sec) Event 0 Simulate steam flow demand to 130%. 2.1 Turbine control valves wide open. 2.28 Vessel water level (L8) trip initiates reactor scram and main turbine and feedwater turbine trips. 2.28 Turbine trip initiates bypass operation to full flow. 2.29 Main turbine stop valves reach 90% open position and initiates recirculation pump trip (RPT). 2.38 Turbine stop valves closed. Turbine bypass valves opening to full flow. 2.4 Recirculation pump motor circuit breakers open causing decrease in case flow to natural circulation. 5.2 Group 1 S/R valves open again to relieve decay heat. 10.2 Group 1 S/R valves close again. 25 Vessel water level reaches L2 setpoint. 28 Low turbine inlet pressure trip initiates main steamline isolation. 33 Main steam isolation valves closed. Bypass valves remain open, exhausting steam in steam-lines downstream of isolation valves. 55 (est) HPCS and RCIC flow enters vessel (not simulated). O 15.1-20
~
[ i ; i
! GESSAR II 22A7007 r i
238 NUCLEAR ISLAND Rev. 7 ( 4 I ' j a ' i Table 15.1-5 SEQUENCE OF EVENTS FOR INADVERTENT SAFETY / RELIEF VALVE OPENING i Time (sec) Event j 0 Initiate opening of 1 S/R valve. i ! 0.5 (est.) Relief flow reaches full flow. ; i , i System establishes new steady-state operation. i 15 (est.) 4 e el 1 1 L I i f t 4 I ; i ! I i i ,I i
- l. I i
f r L I 1 I
+
4
.I J
t
'15.1-21~
am e E ep- Emlwsw-*easme--en-w-<=aww--t=-yewrmeweM+ t-*Ne v w w.1*%w wy re we wwwm .g p w* rv we= yeg = m3 y-> M w =y- eqr etw-y* wpm M gegyqT y7v7WT-' tun-r+'Y v y-wr T- -- Fe'3 T v'
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.1-6 SEQUENCE OF EVENTS FOR, INADVERTENT RHR SHUTDOWN COOLING OPERATION Approximate s Elapsed Time Event 0 Reactor at states B or D (of Appendix 15A) when RHR shutdown cooling inadvertently activated. 0-10 min Slow rise in reactor power.
+10 min Operator may take action to limit power rise.
Flux scram will occur if no action is taken. O O A. N, h a
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 SECTION 15.2 O- TABLES Table Title Page 15.2-1 Sequence of Events for Figure 15.2-1 15.2-59 15.2-2 Sequence of Events for Figure 15.2-2 15.2-60 15.2-3 Sequence of Events for Figure 15.2-3 15.2-61 15.2-4 Sequence of Events for Figure 15.3-4 15.2-62 15.2-5 Sequence of Events for Figure 15.2-5 15.2-63 15.2-6 Sequence of Events for Figure 15.2-6 15.2-64 15.2-7 Post-Transient Release Rate to.the l Containment with Suppression Pool Clernup 15.2-65 15.2-8 Activity Released to the Environment 15.2-66 15.2-9 Estimated ;cs and Atmospheric Dispersion Factors 15.2-67 15.2-10 Egress from Containment Work Area 15.2-68 15.2-11 Design Transient Integrated Egress Doses 15.2-69 15.2-12 Typical Rates of Decay for Condenser O 15.2-13 Vacuum Sequence of Events for Figure 15.2-7 15.2-70 15.2-71 15.2-14 Trip Signals Associated-with Loss of Condenser Vacuum 15.2-72. 15.2-15 Sequence of Events for Figure 15.2-8 15.2-73 15.2-16 Sequence of Events for Figure 15.2-9 15.2-74 15.2-17 Sequence of Events for. Figure 15.2-10 15.2-75~ 15.2-18 Sequence of Events for Failure of RHR Shutdown Cooling 15.2-76 15.2-19 Input Parameters for Evaluation of Failure of RHR Shutdown Cooling 15.2-77 ILLUSTRATIONS Figure Title Page 15.2-1 -Pressure Regulation DownscalelFailure l15.2 15.2-2 Generator Load Rejection, With Bypass'- On. .15.2-80 O
~
15.2-3 Generator-Load Rejection, Without Bypass- 15.2-81
- 15.2-ix
GESSAR II 22A7007 238 NUCLEAR ILLA;4D Rnv. 7 ILLUSTRATIONS (Continued) G Figure Title Page 15.2-4 Turbine Trip With Bypass On, Trip Scram 15.2-82 2
.2-5 Turbine Trip Without Bypass, Trip Scram 15.2-83 15.2-6 MSIV Closure Position Scram 15.2-84 15.2-7 Loss of Condenser Vacuum at 2 in./sec 15.2-85 15.2-8 Loss of Auxiliary Power Transformer 15.2-86 15.2-9 Loss of All Grid Connections 15.2-87 15.2-10 Loss of All Feedwater Flow 15.2-88 15.2-11 ADS /RHR Cooling Loops 15.2-89 15.2-12 Summary of Paths Available to achieve Cold Shutdown 15.2-93 15.2-13 Activity C2 Alternate Shutdown Cooling Path Utilizing RHR Loop A 15.2-94 15.2-14 Activity Cl Alternate Shutdown Cooling Path Utilizing RHR Loop B 15.2-95 15.2-15 RHR Locp B (Suppression Pool Cooling Mode) 15.2-96 15.2-16 RHR Loop C 15.2-97 15.2-17 Vessel Pressure Versus Time 15.2-98 15.2-18 Suppression Pool Temperature Versus Time 15.2-98 O
15.2-x
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 SECTION 15.4 TABLES Table Title Page 15.4-1 Sequence of Events 15.4-25 15.4-2 Rod Block Alarm Distance (BWR/6) 15.4-26 15.4-3 Sequence of Events for Figure 15.4-1 15.4-27 15.4-4 Sequence of Events for Figure 15.4-2 15.4-28 15.4-5 Sequence of Events for Figure 15.4-3 15.4-29 15.4-6 Sequence of Events for the Misplaced Bundle Accident 15.4-30 15.4-7 Input Parameters and Initial Conditions for Fuel Bundle Loading Error 15.4-31 15.4-8 Results of Misplaced Bundle Analysis Equilibrium Cycle 15.4-32 15.4-9 Sequence of Events for Rod Drop Accident 15.4-33 15.4-10 Input Parameters and Initial Conditions For Rod Horth Compliance Calculation 15.4-34 15.4-11 Increment Worth of the Most Reactive Rod Using BPWS 15.4-35 (fs) 15.4-12 Control Rod Drop Accident Evaluation Parameters 15.4-36 15.4-13 Control Rod Drop Accident (Design Basis
' Analysis) Activity Airborne in Condenser 1
(Ci) -15.4-38 15.4-14 Control Rod Drop Accident (Design Basis Analysis) Activity Released to Environment (Ci) 15.4-39 j 15.4-15 Control Rod Drop Accident (Design Basis Analysis) Radiological Effects 15.4-40 f 15.4-16 Control Rod Drop Accident (Realistic l Analysis) Activity Airborne in the ! Condenser (Ci) 15.4-41 _15.4-17 Control Rod-Drop Accident (Realistic
- Analysis) Activity Released to Environment (Ci) -15.4-42 15.4-18 Control Rod Drop Accident (Realistic Analysis) Radiological Effects .15.4-43 i .
l
/~'Y l .LJ l
f 15.4-v/15.4-vi < l
- . . - .- . . . . . . .. ~. - . . . - - - . ..... ~.. - .- - .. - - .. . . - . ~ _..-. . ...- -
4 GESSAR II 22A7007 r 238 NUCLEAR ISLAND Rev. 7 i ! SECTION 15.4 1 1- ILLUSTRATIONS
- i
! Figure Title Page I - j 15.4-1 Startup of Idle Recirculation Loop Pump 15.4-45 ', 15.4-2 Fast Opening of One Main Recirculation Valve
- at 30%/sec 15.4-46 i
15.4-3 Fast Opening of Both Recirculation Valves- ! at ll%/sec 15.4-47 ~ 15.4-4 Leakage Path Model for Rod Drop Accident 15.4-48 i l i B 4 V I 1 i l N l t i 1 1 , .+ 1 j 15.4-vil/15.4-viii-
-....,-.:,..........--.-..:...-.-.,..... -... - ... _ . . - . - .. . - .-. =..- . . - . - - . - . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 15.4.1.2.5 Radiological Consequences 77 evaluation of the radiological consequences is not required for this event, since no radioactive material is released from the fuel. 15.4.2 Rod Withdrawal Error at Power 15.4.2.1 Identification of Causes and Frequency Classification 15.4.2.1.1 Identification of Causes i The Rod Withdrawal Error (RWE) transient results from a procedural error by the operator in which a single control rod or a gang of control rods is withdrawn continuously until the Rod Withdrawal Limiter (RWL) function of the Rod Control and Information System (RCIS) blocks further withdrawal. 15.4.2.1.2 Frequency Classification The frequency of occurrence for the RWE is assumed to be moderate, since definite data do not exist. The frequency of occurrence diminishes as the reactor approaches full power by virtue of the reduced number of control rod movements. A statistical approach, using appropriate conservative acceptance criteria, shows that consequences of the majority of RWEs would be very mild and hardly . noticeable. 15.4.2.2 Sequence of Events and Systems Operation 15.4.2.2.1 Sequence of Events . The sequence of events for this transient is presented in 7-s Table 15.4-1. k_ 15.4-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 15.4.2.2.2 System Operations While operating in the power range in a normal mode of operation, the reactor operator makes a procedural error and withdraws the maximum worth control rod or gang of control rods continuously until the RWL inhibits further withdrawal. The RWL utilizes rod position indications of the selected rod as input. During the course of this event, normal operation of plant instrumentation and conrols is assumed, although no credit is taken for this except as described above. No operation of any engineered safety feature (ESF) is required during this event. 15.4.2.3 Core and System Performance 15.4.2.3.1 Input Parameters and Initial Conditions ] The reactor core is assumed to be on MCPR and MLHGR technical specification limits prior to RWE initiation. A statistical analysis of the rod withdrawal error results (Appendix 15B) initiated from a wide range of operating conditions (exposure, power, flow, rod patterns, xenon conditions, etc) has been per-formed, establishing ailowable rod withdrawal increments appli-cable to all BWR/6 plants. These rod withdrawal increments were determined such that the design basis AMCPR (minimum critical power ratio) for rod withdrawal errors initiated from the techni-cal specification operating limit and mitigated by the RWL system withdrawal restrictions, provides a 95% probability at the 950 confidence level that any randomly occurring RWE will not result in a larger AMCPR. MCPR was verified to be the limiting thermal performance parameter and therefore was uscd to establish the allowable withdrawal increments. The 1% plastic strain limit on the clad was always a less limiting parameter. O 15.4-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 (Oj 15.4.4.2.1.1 Operator Actions The normal sequence of operator actions expected in starting the idle loop is as follows. The operator should: (1) adjust rod pattern, as necessary, for new power level following idle loop start; (2) determine that the idle recirculation pump suction and discharge block values are open and that the flow control valve in the idle loop is at minimum position and, if not, place them in this configuration; (3) readjust flow of the running loop downward to less than half of the rated flow; (4) determine that the temperature difference between the [) v two loops is no more than 50*F; (5) start the idle loop pump and adjust flow to match the adjacent loop flow (monitor reactor power) ; and (6) readjust power, as necessary, to satisfy plant require- , ments per standard procedure. I l NOTE: The time to do the above work is approximately l _ [ 1/2 hour. 15.4.4.2.2 Systems Operation This event assumes and takes credit for normal functioning of plant instrumentation and controls. tk) protection systems action is anticipated. No ESF action occurs as a result of the transient. o i I 15.4-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. G
~
15.4.4.3 Core and System Performance 15.4.4.3.1 Input Parameters and Initial Conditions , One recirculation loop is idle and filled with cold water (100*F). (Normal procedure when starting an idle loop with one pump already running requires that the indicated idle loop temperature be no more than 50 F lower than the indicated active loop temperature.) The active recirculation loop is operating with the flow control valve position that produces about 70% of normal rated jet pump diffuser flow in the active jet pumps. The core is receiving 33% of its normal rated flow. The remainder of the coolant flows in the reverse direction through the inactive jet pumps. The idle recirculation pump suction and discharge block valves are open and the recirculation flow control valve is closed to its minimum open position. (Normal procedure requires leaving an idle loop in this condition to maintain the loop temperature within the required limits for restart.) 15.4.4.3.2 Results ] The transient response to the incorrect startup of a cold, idle recirculation loop is shown in Figure 15.4-1. Shortly after the pump begins to move, a surce in flow from the started jet pump diffusers causes the core inlet flow to rise sharply. The motor approaches synchronous speed in approximately 3 see because of the assumed minimum pump and motor inertia. A short-duration neutron flux peak is produced as the colder, increasing core flow reduces the void volume. Surface heat flux follows the slower response of the fuel and peaks at 80% of rated 15.4-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 () 15.4.4.3.2 Results (Continued)
~
before decreasing after the cold water washed out of the loop at about 18 sec. No damage occurs to the fuel barrier and MCPR remains significantly above the safety limit as the reactor settles out at its new steady-state condition. Therefore, this event does not have to reanalyzed for specific core configurations. _ 15.4.4.4 Barrier Performance No evaluation of barrier performance is required for this event since no significant pressure increases are incurred during this transient (Figure 15.4-1). 15.4.4.5 Radiological Consequences () An evaluation of the radiological consequences is not required for this event, since no radioactive material is released from the fuel. l j 15.4.5 Recirculation Flow Control Failure with Increasing Flow-15.4.5.1 Identification of Causes and Frequency Classification 15.4.5.1.1 Identification of Causes Failure of the master controller of neutron flux controller can cause an increase in the core coolant flow rate. Failure within a loop's flow controller can also cause an increase in core coolant flow rate. 15.4.5.1.2 Frequency Classification () This transient disturbance is classified as an incident of moderate frequency. 15.4-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 15.4.5.2 Sequence of Events and Systems Operation 15.4.5.2.1 Sequence of Events 15.4.5.2.1.1 Fast Opening of One Recirculation Valve Table 15.4-4 lists the sequence of events for Figure 15.4-2. 15.4.5.2.1.2 Fast Opening of Two Recirculation Valves Table 15.4-5 lists the sequence of events for Figure 15.4-3. 15.4.5.2.1.3 Identification of Operator Actions Initial action by the operator should include: (1) transfer flow control to manual and reduce flow to minimum, and (2) identify cause of failure. Reactor pressure will be controlled as required, depending on whether a restart or cooldown is planned. In general, the corrective action would be to hold reactor pressure and condenser vacuum for restart after the malfunctioning flow controller has been repaired. The following is the sequence of operator actions expected during the course of the event, assuming restart. The operator should: (1) observe that all rods are in; (2) check the reactor water level and maintain above low level (L2) trip to prevent MSLIVs from isolating; (3) switch the reactor mode switch to the STARTUP position; 15.4-12
i CESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 () 15.4.5.4 Barrier Performance 15.4.5.4.1 Fast Opening of One Recirculation Valve This transient results in a very slight increase in reactor vessel pressure (Figure 15.4-2) and therefore represents no threat to the RCPB. .; 15.4.5.4.2 Fast Opening of Two Recirculation Valves This transient results in a very slight increase in reactor vessel pressure (Figure 15.4-3) and therefore represents no threat to the RCPB. 15.4.5.5 Radiological Consequences An evaluation of the radiological consequences is not required for () this event, since no radioactive material is released from the fuel. 15.4.6 Chemical and Volume Control System Malfunctions Not applicable to BWRs. This is a PWR event. 15.4.7 Misplaced Bundle Accident , 15.4.7.1 Identification of Causes and Frequency Classification 15.4.7.1.1 Identification of Causes-The event discussed in this section'is the improper loading of a i fuel bundle and subsequent operation of the core. Three errors must occur for this event to take place in the equilibrium core loading. First, a bundle must be misloaded into a wrong location Second, the bundle which was supposed to be loaded () in the core. where the mislocation occurred would have to also be put in an 15.4-15
GESSAR II 22A7007 238 WUCLEAR ISLAND Rev. 7 15.4.7.1.1 Identification of Causes (Continued) incorrect location or discharged. Third, the misplaced bundles would have to be overlooked during the core verification process performed following core loading. 15.4.7.1.2 Frequency Classification This unlikely event occurs when a fuel bundle is loaded into the wrong location in the core. It is assumed the bundle is misplaced to the worst possible location, and the plant is operated with the mislocated bundle. This event is categorized as an infrequency incident based on the following data: Expected Frequency: 0.002 events / operating cycle The above number is based upon past experience. 15.4.7.2 Sequence of Events and Systems Operation O 15.4.7.2.1 Sequence of Events The postulated sequence of events for the misplaced bundle accident (MBA) is presented in Table 15.4-6. 15.4.7.2.2 Systems Operation A fuel loading error, undetected by in-core instrumentation follow-ing fueling operations, ma-j result in an undetected reduction in thermal margin during power operations. For the analysis reported herein, no credit for detection is taken and, therefore, no corrective operator action or automatic protection system functioning is assumed to occur. O 15.4-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O - Table 15.4-1 SEQUENCE OF EVENTS Elapsed Time (sec) O Core is operated in a typical control rod pattern on limits 0 Operator withdraws a single rod or gang of rods continuously 1 The local power in the vicinity of the withdrawn rod (or gang) increases. Gross core power increases.
) %4 RWL blocks further withdrawal N25 Core stabilizes at slightly higher core power level For.a 1.0 ft RWL' incremental withdrawal block. Time would be longer for a larger block, since rods are with-drawn at approximately 3 in./sec.-
O 15.4-25 l
' "^ .-, - .-. --. , . _ _ , , . _, ._
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.4-2 ROD BLOCK ALARM DISTANCES (BWR/6) Power Range (% of rated) Allowable Withdrawal Distance (ft) 60 - 100 1.0 20 - 70 2.0 0- 20 no restrictions
- The BPWS function of the RCIS provides control of rod withdrawals below the 20% power setpoint and allows a maximum withdrawal distance of 9 ft.
O O 15.4-26
I i GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 7 i l-l Table 15.4-3 i
SEQUENCE OF EVENTS FOR FIGURE 15.4-1 4 Time (sec) Event , O Start pump motor O.30 Jet pump diffuser flows on started pump side become positive l , 3.0 Pump motor at full speed and drive flow at about 21% of f rated l l 18.0 Last of cold water leaves recirculation drive loop
- (est)
I 18.1 Peak value of core inlet subcooling 50 Reactor variables settle into new steady state l iO 1 x i 15.4-27 ,
. . - _ _ . , . - . _..._._...._._.a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 , I Table 15.4-4 O, SEQUENCE OF EVENTS FOR FIGURE 15.4-2 Time (sec) Event 0 Simulate failure of single loop control 1.3 Reactor APRM high-flux scram trip initiated 3.0 Turbine control valves start to close upon falling (est) turbine pressure 6.5 Recirculation pump drive motors trip due to L3 25 Turbine control valves closed. Turbine pressure below pressure regulator setpoints
>100 Reactor variables settle into new steady-state (est)
O O 15.4-28
GESSAR II 22A/UU7 238 NUCLEAR ISLAND Rev. 7
) Table 15.4-5 SEQUENCE OF EVENTS FOR FIGURE 15.4-3 Time (sec) Event 0 Initiate failure of master controller 1.6 Reactor APRM high-flux scram trip initiated l
3.5 Turbine control valves start to close upon falling (est) turbine pressure 5.6 Recirculation pump drive motors trip due to L3 32.0 Turbine control valves closed. Turbine pressure below (est) pressure regulator setpoints i'
>100 Reactor variables settle into new steady-state (est)
'O !O 15.4-29 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.4-6 SEQUENCE OF EVENTS FOR THE MISPLACED BUNDLE ACCIDENT (1) During the core loading operation, a bundle is loaded into the wrong core location. (2) Subsequently, the bundle designated for this location is incorrectly loaded into the location of the previous bundle. (3) During the core verification procedure, the two errors are not observed. (4) The plant is brought to full power operation without detecting misplaced bundle. (5) The plant continues to operate throughout the cycle. O O 15.4-30
1 GESSAR II 22A7007 238 NUCLEAR ISLAND ~Rev. 6 O Table 15.4-7 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TIIE FUEL BUNDLE LOADING ERROR (1) Power (% rated) 100 (2) Flow (% rated) 106 (3) MCPR operating limit
- 1.20 (4) MLHGR operating limit (kW/ft)* 13.4 j (5) Core Exposure End of Cycle
*These are above the current operating limits. Since these -
limits do not go into the calculation of the MCPR associated with a mislocated bundle, differences in the safety operating limits will not effect these results. _ O ) n m i 15.4-31
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.4-8 l RESULTS OF MISPLACED BUNDLE ANALYSIS EQUILIBRIUM CYCLE (1) MCPR Safety Limit 1.07 (2) MCPR with misplaced bundle 1.14 (3) LHGR 1% plastic strain limit >20 kW/ft (4) LHGR with misplaced bundle
- 14.9 Does not include any densification penalty.
O O 15.4-32
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.4-9 SEQUENCE OF EVENTS FOR ROD DROP ACCIDENT Approximate Elapsed Time (sec) Event Reactor is operating at 50% rod density pattern. Maximum worth control rod blade becomes decoupled from the CRD. Operator selects and withdraws the control rod drive of the decoupled rod either individually or , along with other control rods assigned to the RCIS group. Decoupled control rod sticks in the fully inserted or an intermediate bank position. O Control rod becomes unstuck and drops to the drive position at the nominal measured velocity plus O <1 three standard deviations. Reactor goes on a positive period and the initial power increase is terminated by the doppler coefficient.
<1 APRM 120% power signal scrams reactor.
<5 Scram terminates accident.
O 15.4-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 15.4-10 h INPUT PARAMETERS AND INITIAL CONDITIONS FOR ROD WORTH COMPLIANCE CALCULATION
- 1. Reactor Power (% rated) 1
- 2. Reactor Flow (% rated) 100
- 3. Core Average Exposure (mwd /t) Most reac{lve Point in cycle
- 4. Control Rod Fraction %0.50
- 5. Average Fuel Temperature (*C) 286
- 6. Average Moderator Temperature (*C) 286
- 7. Xenon State None
- 8. Core Average Void Fraction (%) 0 0
O 15.4-34
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O Table 15.4-11 INCREMENT WORTH OF THE MOST REACTIVE ROD USING BPWS Control Banked Control Core Rod At Rod Drops Increase Condition Group Notch (I,J) From-To (keff) 3000 7 04 (26,35) 00-08 0.00248 3000 7 08 (26,35) 00-12 0.00278 3000 7 12 (26,35) 00-48 0.00269 3000 7 48 (26,35) 00-48 0.00198 NOTE: The following assumptions were made to ensure that the rod worths were conservatively high for the BPWS: (a) BOC (b) Hot Startup (c) No Xenon O O 15.4-35
s GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 , T Table 15.4-12 g CONTROL ROD DROP ACCIDENT EVALUATION PARAMETERS ,- Design Realistic Basis Basis , Assumptions Assumptions I. Data and assumptions used to estimate radioactive source from postulated 4 accidents: A. Power level 3651 MWt 3651 MWt B. Burnup NA NA -, C. Fuel damaged 770 rods 770 rods - D. Release of activity by nuclide Table 15.4-14 Table-15.4-17 E. Iodine fractions: (1) Organic 0 0 (2) Elemental 1 1 (3) Particulate 0 0 F. Reactor coolant activity before NA NA the accident G. Peaking factor 1.5 1.0 II. Data and assumptions used to estimate activity released: A. Condenser leak rate (%/ day) 1.0 0.5 _. B. Turbine building leak rate (%/ day) NA 1327. , C. Valve closure time (sec) NA 5 D. Absorption and filtration efficiencies: ,, (1) Organic iodine NA NA (2) Elemental iodine NA NA s (3) Particulate iodine NA NA x , g (4) Particulate fission products NA NA E. Recirculation system parameters: (1) Flow rate NA NA (2) Mixing efficiency NA NA (3) Filter efficiency NA NA F. Containment spray parameters NA NA s (flow rate, drop size, etc) . s G. Containment volumes NA NA H. All other pertinent data and None None .. s assumptions . 15.4-36 - -
i' GESSAR II ',' 22A7007 , 238 NUCLEAR ISLAND
- Rev. 7 i s
~
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' Table i5.4-12 (Continued) s -
m ,- Design Realistic ' Basis Basis
'4 Assumptions Assumptions III. Dispersion Data: ,A
{ A. Site B'undary'and PZ distances,;(m) *
- a l B. X/Q's for' time intervals of: >
i (1) 0-1 hr - SB/LPZ -' 2.0E-3/1.0E-2 2.0E-3/1.0E-3 (2) 1-8 hr - SB/LPZ , , - 3.8E-4 3.8E-4 (3) 8-16 hr -~SB/LPZ l.0E-4 1.0E-4 i (4) 16 hr-3 days LPZ 3.4E-5 3.4E-5 (5) 3-26 day - LPZ - 7.5E-6_ 7.5E-6 IV. Dose Data: A. Method of dose calculatlon ' Reference Reference 3 . 4-B. Dose conversion assumptions Reference Reference 3 4 C. Peak activity, concentrations Table 15.'4-13' Table 15.4-16 , in condenser i D. Doses . Table 15.4-15~ Table 15.4-18 j 1 3 .;
~~
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Table 15.4-13 CONTROL ROD DROP ACCIDENT (DESIGN BASIS ANALYSIS) ACTIVITY AIRBORNE IN CONDENSER (Ci) Isotope 1 min 30 min 1 hr 2 hr 4 hr 8 hr 12 hr 1 day 4 day 30 day Il31 2.2E 03 2.2E 03 2.2E 03 2.2E 03 2.2E C3 2.lE 03 2.lE 03 2.0E 03 1.5E 03 1.2E O2 I132 3.6E 03 3.lE 03 2.7E 03 2.OE 03 1.lE 03 3.2E O2 9.4E 01 2.5E 00 7.5E-lO O. Il33 3.3E 03 3.3E 03 3.2E 03 3.lE 03 2.9E 03 2.6E 03 2.2E 03 1.5E 03 1.3E O2 9.5E-08 1134 5.6E 03 3.8E 03 2.6E 03 1.2E 03 2.4E O2 1.0E 01 4.2E-01 3.lE-05 d. O. I135 4.7E 03 4.5E 03 4.2E 03 3.8E 03 3.lE 03 2.OE 03 1.3E 03 3.7E O2 1.8E-Ol O. Total I 1.9E 04 1.7E 04 1.5E 04 1.2E 04 9.5E 03 7.UE 03 5.7s 03 3.9E 03 1.6E 03 1.2E O2 Kr83m 2.5E 04 2.lE 04 1.8E 04 1.2E 04 5.7E 03 1.3E 03 2.8E 02 3.2E 00 5.8E-12 O. $ ? Kr85m 6.lE 04 5.6E 04 5.2E 04 4.5E 04 3.3E 04 1.8E 04 9.5E 03 1.5E 03 2.OE-02 0. M p "y' Kr85 1.6E 03 1.6E 03 1.6E 03 1.6E 03 1.6E 03 1.6E 03 1.6E 03 1.5E 03 1.5E 03 1.2E 03 HH 1.8E 02 2.5E 01 01 U3 H Kr87 1.2E 05 9.5E 04 7.3E 04 4.2E 04 1.4E 04 1.6E 03 0. 7,8E-06 b Kr88 1.8E 05 1.6E 05 1.4E 05 1.lE 05 6.6E 04 2.4E 04 9.lE 03 -4.6E O2 0. 2 0 Kr89 1.8E 05 3.lE 03 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.4E 03 1.2E 03 2.OE O2 Xel31m 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.5E 03 1.4E 03 1.2E 03 2.OE O2 Xel33m 6.lE 04 6.lE 04 6.OE 04 6.OE 04 5.8E 04 5.5E 04 5.2E 04 4.4E 04 1.7E 04 4.lE 00 Xel33 3.6E 05 3.5E 05 3.5E 05 3.5E 05 3.5E 05 3.4E 05 3.3E 05 3.lE 05 2.OE 05 5.lE 03 Xel35m 9.7E 04 2.6E 04 6.7E 03 4.4E O2 1.9E 00 3.6E-05 6.9E-lO '9 . O. O. Xel35 6.5E 04 6.2E 04 6.OE 04 5.6E 04 4.8E 04 3.5E 04 2.6E 04 1.OE 04 4.3E 01 O. Xel37 3.9E 05 2.lE 03 9.2E 00 1.8E-04 7.OE-14 O. O. O. O. O. Xel38 4.3E 05 1.nE 05 2.4E 04 1.3E 03 3.6E 00 2.9E-05 2.4E-10 O. O. 01 m I$ o >*
<4 Total NG 2.OE 06 9.4E 05 7.9E 05 6.8E 05 5.7E 05 4.8E 05 4.3E 05 3.7E 05 2.2E 05 6.5E 03
- Cj
-a -a O O O
x i LJ L/ Table 15.4-14 CONTROL ROD DROP ACCIDENT (DESIGN BASIS ANALYSIS) ACTIVITY RELEASED TO ENVIRONMENT (Ci) Isotope 1 min 30 min 1 hr 2 hr 4 hr 8 hr 12 hr 1 day 4 day 30 day Il31 1.5E-02 4.6E-Ol 9.lE-01 1.8E 00 3.6E 00 7.2E 00 1.lE 01 2.lE 01 7.3E 01 2.lE O2 I132 2.5E-02 7.OE-01 1.3E 00 2.3E 00 3.5E 00 4.5E 00 4.8E 00 4.9E 00 4.9E 00 4.9E 00 Il33 2.3E-02 6.9E-01 1.4E 00 2.7E 00 5.2E 00 9.8E 00 1.4E 01 2.3E 01 4.OE 01 4.lE 01 I134 3.9E-02 0.8E-01 1.6E 00 2.4E 00 2.9E 00 3.0E 00 3.0E 00 3.OE 00 3.OE 00 3.OE 00 Il35 3.3E-02 9.5E-Ol 1.9E 00 3.5E 00 6.4E 00 1.lE 01 1.3E 01 1.7E 01 1.9E 01 1.9E 01 w w 4.6E 01 1.4E O2 2.8E O2
- Total I 1.4E-Ol 3.8E 00 7.lE 00 1.3E 01 2.2E 01 3.5E 01 6.9E 01 z
g Kr83m 1.8E-Ol 4.9E 00 8.9E 00 1.5E 01 2.2E 01 2.7E 01 2.8E 01 2.8E 01 2.8E 01 2.8E 01 $$ (n Kr85m 4.2E-01 1.2E 01 2.4E 01 4.4E 01 2.7E 00 1.2E 00 1.4E O2 1.6E O2 1.6E O2 1.6E O2 {my b Kr85 1.lE-02 3.3E-Ol 6.5E-01 1.3E 00 2.6E 00 5.2E 00 7,8E 00 1.6E 01 6.lE 01 4.OE O2 W HH 9.5E 01 mH Kr87 8.7E-01 2.3E 01 4.0E 01 6.4E 01 8.5E 01 9.4E 01 9.5E 01 9.5E 01 9.5E 01 3.OE O2 3.OE O2 b z Kr88 1.2E 00 3.5E 01 6.6E 01 1.2E O2 1.9E O2 2.6E O2 2.8E O2 3.OE O2 O Kr89 1.4E 00 7.lE 00 7.1E 00 7.lE 00 7.lE 00 7.lE 00 7.lE 00 7.lE 00 7.lE 00 7.lE 00 Xel31m 1.lE-02 3.2E-Ol 6.3E-OL 1.3E 00 2.5E 00 5.OE 00 7.5E 00 1.5E 01 5.4E 01 2.OE O2 Xel33m 4.3E-01 1.3E 01 2.5E 01 5.OE 01 9.9E 01 1.9E 02 2.8E O2 5.2E O2 1.4E 03 1.9E 03 Xel33 2.5E 00 7.4E 01 1.5E O2 2.9E O2 5.9E O2 1.2E 03 1.7E 03 3.3E 03 1.lE 04 2.5E 04 Xel35m 6.9E -01 1.2E 01 1.4E 01 1.5E 01 1.5E 01 1.5E 01 1.5E 01 1.5E 01 1.5E 01 1.5E 01 Xel35 4.5E-Ol 1.3E 01 2.6E 01 5.OE 01 9.3E 01 1.6E O2 2.lE O2 3.0E O2 3.5E O2 3.5E O2 Xel37 3.OE 00 1.8E 01 1.8E 01 1.8E 01 1.8E 01 1.8E 01 1.8E 01 1.8E 01 1.8E 01 1.8E 01 Xel38 3.OE 00 4.9E 01 6.OE 01 6.3E 01 6.4E 01 6.4E 01 6.4E 01 6.4E 01 6.4E 01 6.4E 01 mN o :p
< -J Total NG 1.4E 01 2.6E O2 4.4E O2 7.4E O2 1.3E 03 2.lE 03 2.9E 03 4.9E 03 1.3E 04 2.8E 04 - o o
-J -J
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 TABLE 15.4-15 CONTROL ROD DROP ACCIDENT (DESIGN BASIS ANALYSIS) Radiological Effects Whole Body Inhalation Dose (rem) Dose (rem) Exclusion Area 0.22 2.55 j Low Population Zone 0.16 4.08f 9 O 15.4-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 51.4-16 CONTROL ROD DROP ACCIDENT (REALISTIC ANALYSIS) ACTIVITY AIRBORNE IN THE CONDENSER (Ci) Isotope 1 min 1 hr 2 hrs 8 hrs 1 day 4 days 30 days I131 2.92E-01 2.91E-01 2.90E-01 2.84E-01 2.68E-01 2.07E-01 2.20E-02 1 Il32 4.47E-02 3.31E-02 2.45E-02 3.96E-03 3,08E-05 9.65c-15 0. Il33 1.43E-01 1.38E-01 1.34E-01 1.09E-01 6 42E-02 5.80E-03 5.46E-12 Il34 3.36E-02 1.54E-02 6.99E-03 6.04E-05 1.89E-10 0. O. Il35 1.09E-01 9.84E-02 8.86E-02 4.72E-02 8.77E-03 4.48E-06 0. Total 6.23E-01 5.77E*01 5.44E-01 4.45E-01 3.41E-01 2.13E-01 2.20E-02 ! Kr83m 3.35E 01 2.32E 01 1.59E 01 1.68E 00 4.20E-03 7.78E-15 O. Kr85m 2.28E 02 1.95E 02 1.67E 02 6.60E 01 5.53E 00 7.73E-05 O. KrG5 2.26E 02 2.26E 02 2.26E 02 2.26E 02 2.25E 02 2.21E 02 1.94E 02 Kr87 1.91E 02 1.12E 02 6.46E 01 2.42E 00 3.80E-04 0. O. Kr88 4.30E 02 3.37E 02 2.63E 02 5.94E 01 1.13E 00 1.94E-08 0. Kr89 1.13E-01 2.67E-07 5.06E-13 0. O. O. 0.. Xel31m 2.87E 01 2.86E 01 2.85E 01 2.81E 01 2.69E 01 2.23E 01 4.36E 00 Xel33m 4.41E 02 4.35E 02 4.29E 02 3.97E 02 3.21E 02 1.24E 02 ~3.39E-02 Xel33 4.27E 03 4.25E 03 4.22E 03 4.08E 03 3.73E 03 2.47E 03 7.17E 01 Xel35m 3.23E 00 2.23E-Cl 1.47E-02 1.22E-09 9. O. O. Xel35 8.99E 02 8.34E C2 7.73E 02 4.91E 02 1.46E 02 6.17E-01 0. Xel37 4. ole-01 9.45E-06 1.86E-10 0. O. O. O. Xel38 6.17E 01 3.46E 00 1.84E-01 4.26E-09 0. O. O. Total 6.81E 03 6.44E 03 6.19E 03 5.35E 03 4.45E 03 2.84E 03 2.70E 02 l
-15.4-41
b GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 7 Table 15.4-17 CONTROL ROD DROP ACCIDENT (REALISTIC ANALYSIS) ACTIVITY RELEASED TO TIIE ENVIRONMENT (Ci) Isotope 1 min 1 hr 2 hrs 8 hrs 1 day 4 days 30 days Il31 4.66E-09 1.41E-05 4.79E-05 3.72E-04 1.29E-03 4.84E-03 1.55E-02 I132 7.14E-10 1.78E-06 5.07E-06 1.72E-05 1.99E-05 1.99E-05 1.99E-05 Il33 2.28E-09 6.75E-06 2.25E-05 1.59E-04 4.41E-04 8.05E-04 8.42E-04 Il34 5.39E-10 1.00E-06 2.21E-06 3.67E-06 3.69E-06 3.69E-06 3.69E-06 I135 1.74E-09 4.93E-06 1.58E-05 8.88E-05 1.65E-04 1.82E-04 1.82E-04 Total 9.94E-09 2.86E-05 9.34E-05 6.41E-04 1.92E-03 5.85E-03 1.66E-02 Kr83m 5.35E-07 1.28E-03 3.50E-03 1.02E-02 1.12E-02 1.12E-02 1.12E-02 Kr85m 3.63E-06 9.97E-03 3.09E-02 1.51E-01 2.32E-01 2.40E-01 2.40E-01 Kr85 3.61E-06 1.09E-02 3.72E-02 2.92E-01 1.04E 00 4.40E 00 3.13E 01 Kr87 3.06E-06 6.59E-03 1.64E-02 3.60E-02 3.69E-02 3.69E-02 3.69E-02 Kr88 6.86E-06 1.78E-02 5.21E-02 2.01E-01 2.50E-01 2.50E-01 2.50E-01 Kr89 1.94E-09 8.96E-08 8.96E-08 8.96E-08 8.96E-08 8.96E-08 8.96E-08 Xel31m 4.57E-07 1.38E-03 4.70E-03 3.67E-02 1.28E-01 4.97E-01 1.93E 00 Xel33m 7.03E-06 2.llE-02 7.14E-02 5.37E-01 1.73E 00 4.85E 00 6.82E 00 Xe133 6.81E-05 2.06E-01 6.98E-01 5.39E 00 1.84E 01 6.43E 01 1.52E 02 Xe135m 5.23E-08 3.49E-05 4.30E-05 4.38E-05 4.38E-05 4.38E-05 4.38E-05 Xel35 1.43E-05 4.14E-02 1.35E-01 8.29E-01 1.77E 00 2.17E 00 2.18E 00 Xel37 6.80E-09 4.48E-07 4.48E-07 4.48E-07 4.48E-07 4.48E-07 4.48E-07 Xc138 1.00E-06 6.04E-04 7.21E-04 7.31E-04 7.31E-04 7.31E-04 7.31E-04 Total 1.09E-04 3.17E-01 1.05E 00 7.48E 00 2.36E 01 7.68E 01 1.95E 02 O 15.4-42
l GESSAR II 22A7007 i ' 238 NUCLEAR ISLAND Rev. 7 i l l
- TABLE 15.4-18 l l CONTROL ROD DROP ACCIDENT (REALISTIC ANALYSIS) 4 i RADIOLOGICAL EFFECTS .
t I ' i 4 1 4 Whole Body Inhalation l, Dose (rem) Dose (rem) ( c 1 4 . i 1 Exclucion Area '9.4E-05 5.4E-05 ' i o ! i
- Low Population Zone 1.7E-04 2.0E-04 i
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- 1. DESIGN 8 ASIS EVALUATION CONDENSER L ENVIRONMENT
- 2. REALISTIC BASIS EVALUATION ENVIRONMENT CONDENSER TURBINE BUILDING L
O Figure 15.4-4. Leakage Path Model for Rod Drop Accident O 15.4-48
_ -- . _ - _ _ = . _ . . - - . - - .- - GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2 BWR PREVENTION AND MITICATION CAPABILITY f ) This section describes the basic features and capabilities of the BWR/6 238 Nuclear Island which prevent or mitigate
- severe accidents.
The 238 Nuclear Island includes design features which prevent ! damage to the reactor core from transients or accidents (Subsection 15D.2.1). In the extremely unlikely event that extensive multiple failures result in core damage, additional features are provided to mitigate the effects of those accidents (Subsection 15D.2.2). The adequacy of this design is quantified by analysis of the probability of. core damage and the risk to the public. These evalua.tions are discussed in the probabilistic risk assessment presented in Section 15D.3. () In addition to the design features and capabilities, the development of procedural guidelines to be used by operators ; in the event of an accident are made part of the standard design process. A discussion of the emergency procedure I guidelines is provided in Subsection 15D.2.3. 15D.2.1 BWR Prevention Features and Capabilities I The BWR safety approach has traditionally. stressed the prevention of core damage as a key objective of the design to ensure plant safety. By providing multiple and diverse methods of water injection to the reactor vessel, the like-lihood of core damage due to inadequate core cooling is ; minimized. Subsection 15D.3.3, Probability of Core Damage, provides quantitative results to supportLthe high degree of success of this' safety approach. . O 230-Al 15D.2-1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.1 BWR Prevention Features and Capabilities (Continued) The basic features and capability of the 238 Nuclear Island are described in GESSAR Chapters 5, 6 and 7. Following the accident at Three Mile Island, several design changes were identified by the NRC and the Nuclear Industry which resulted in relatively minor changes to the basic design described in Chapters 5, 6 and 7. Some of these changes (Subsection 15D.2.1.3) have an effect on the probabilistic risk assess-ment discussed in Section 15D.3. Other changes, though not quantifiable in terms suitable for inclusion in a PRA, provide an additional positive improvement to plant safety. ] Changes in response to all applicable post-TMI requirements are described in Appendix 1A. The principal BWR core protection functions are provided by
- 1) the reactor protection system, 2) the systems to supply wat.er to the reactor core to provide adequate care cooling and 3) the systems to remove the decay heat. These functions are described below.
15D.2.1.1 Reactor Protection The reactor protection system is described in Chapter 7 (Section 7.2). The reactor protection function is performed by a control rod drive system (Section 4.6) and the Standby Liquid Control System (Section 7.4). These systems provide reliable and diverse methods of controlling reactor neutron flux and achieving reactivity control when protective action is required. 15D.2.1.2 Core Cooling The ability of the BWR plant to supply water to the core following transients and accidents is provided by a combi-nation of high pressure and low pressure water delivery 15D.2-2 230-Cl
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 () 15D.2.1.2 Core Cooling (Continued) systems as listed in Figure 15D.2-1. The means to rapidly depressurize the reactor vessel in the event that high pressure systems are unavailable is also provided. The high pressure systems consist of the main feedwater system, the High Pressure Core Spray (HPCS) system (Section 6.3), the Reactor Core Isolation Cooling (RCIC) system (Subsection 5.4.6) and flow from the Control Rod Drive (CRD) Hydraulic system. The low pressure injection systems consist of three inde- i t pendent low pressure coolant injection (LPCI) loops which { are part of the Residual Heat Removal (RHR) system, the Low Pressure Core Spray system (LPCS), and the condensate system. , The LPCI and LPCS systems are described in Section 6.3. The condensate system can provide makeup water to the reactor at
~
low pressure independent of the availability of the feedwater system. The capability to rapidly depressurize'the pressure vessel is provided by the Automatic Depressurization System (ADS) which is described .n Section 7.3. In addition to this automatic system, the operator can manually depressurize_the i pressure vessel through the Safety Relief Valves or by using l the main condenser as a heat sink. , I i i
- Several key protection functions result from the ability to i depressurize the BWR. If the high pressure makeup _ systems should all be unavailable, actuation of the ADS will depres-surize the reactor vessel so that one or more of the low pressure systems may be used to maintain water level. When depressurized, r.11 water sources are available for injection into the reactor vessel. These sources include the suppression
[ 15D.2-3 230-C2
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.1.2 Core Cooling (Continued) h pool, the condensate storage tank, main condensers and the service water supply. Depressurization also causes an increase in the actual water level in the core through creation of voids. Finally, RPV depressurization reduces any dynamic loads which the containment may be experiencing. These features provide a unidirectional operator action (depressurization) in response to core inventory threatening situations which facilitates the development of the symptom- ] based emergency procedure guidelines. (Subsection 15D.2.3) Unidirectional operator action in response to events is a key feature of the BWR design. Although water delivery and depressurization systems perform automatically, the operator is not burdened with decisions as to the correct course of action to assure adequate core cooling. In general, these actions are to maintain RPV water level in the normal range, and, in the extreme, depressurize the reactor vessel and ensure at least one water delivery system is operating. Subsection 15D.2.3 provides further detail on the development and content of the Emergency Procedure Guidelines. The diversity of water delivery systems is a key feature of l the BWR design which assures prevention of core damage l following accidents or transients. Diversity is also found on the power supplies for the equipment, the motive force for components and in the method of water delivery. l Diversity in power supply to High Pressure and Low Pressure j ECCS equipment is provided through use of different diesel l generator vendors for the division 3 (H2CS) power as compared with di'rision 1 or 2 (LPCI, LPCS) power. Direct-current powered valves and controls are supplied by batteries and alternating-current powered inverters. The motive forces used to drive high pressure water delivery systems are 15D.2-4 230-C3
{ GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 () 15D.2.1.2 Core Cooling (Continued) diverse through use of a steam turbine for RCIC and Diesel Electric drive for HPCS. Water delivery is provided through flooding (delivery to the core shroud with flow up through the bottom of the core) and spray (delivery through spray nozzles above the top of active fuel). These diverse features of the system design add to the reliability of the systems and contribute to the low likelihood of core damage. The number of pumps and pump flow capability associated with high pressure and low pressure injection systems are summa-rized in Table 15D.2-1. The plant response to design basis accidents is provided in Chapter 15. These analyses only take credit for the emergency core cooling system (ECCS). , The Chapter 15 accident analyses show that the plant response O is well within the licensing basis of the plant. The benefit of the BWR single vessel design is that at least seven pumps ] in addition to the ECCS and RCIC pumps have the capability to inject directly to the reactor vessel and maintain the water level even if none of the ECCS pumps are available. i The probabilistic risk assessment in Subsection 15D.3.3 provides success criteria for adequate core cooling during transients and accidents which have degraded far beyond l i design basis assumptions. These criteria are based on realistic calculations of not exceeding 2200*F peak clad temperature. These criteria show the capability of the high l pressure and low pressure water delivery systems to provide i l adequate core cooling for all initiating events including l the large loss-of-coolant accident. 1 Indication of successful core cooling is provided to the (~'
\ operators in the control room by the reactor pressure vessel I 15D.2-5 230-C4 i
l
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.1.2 Core Cooling (Continued) the measurement concept) . This system, which is described further in Section 7.7, consists of redundant condensing chambers and differential pressure transmitterc. By pro-viding an indication of water level inventory in the vessel above the reactor core, and by providing an indication of the trend of the water level, the operator is assured that water is available to adequately cool the core. As backup to the trend indication of water level, flow indication of the high pressure and low pressure systems which supply water to the core is also provided. If there is an indication that the water level is dropping below the normal range, the operator has a single clear and direct action to assure continued adequate core cooling: increase the water level. 15D.2.1.3 Decay Heat Removal 15D.2.1.3.1 Vessel Heat Removal Removal of the reactor decay heat from the pressure vessel is accomplished via release through the main steam system to the main condenser, through the RHR heat exchangers in either the steam condensing mode or shutdown cooling mode, or through Safety / Relief valves to the suppression pool. These heat exchangers also are capable of removing the energy stored in the suppression pool. During an accident or transient, if the main steam line is isolated, or if a momentary pressure increase occurs, decay heat removal from the core is accomplished by discharge of steam to the suppression pool through Safety / Relief valves (shown simplistically in Figure 15D.2-2). During relief valve action, the strong natural circulation down the core shroud and up through the core and steam separators ensures a passive means of decay heat removal within the reactor vessel with no reliance on active systems external to the 230-A6 15D.2-6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 4 ( 15D.2.1.3.1 Vessel Heat Removal (Continued) reactor vessel. Accumulation of non-condensible gases is
}
l mir.imized by relief valve operation. Further, non condensi-ble gases in the top of the vessel do not affect the natural l i circulation flow. i i i ! If the main steam line is not isolated, core decay heat l removal can be achieved via the Main Steam System provided a j condenser vacuum can be maintained and a return flow path to l the reactor vessel can be established through the condensate pumps, condensate booster pumps and feedwater pum9s. This mode of operation requires the availability of offsite power, but is the normal means of decay heat removal. Decay heat is removed from the core by natural circulation as previously described for conditions of main steam line isolation. Finally, when the reactor has been depressurized below 150 psia, the shutdown cooling mode of the RHE system may be used (Subsection 5.4.7). This system is operable on either onsite or offsite power and provides long-term core cooling by providing flow directly to and from the reactor vessel. The suppression pool acts ac a passive heat sink to absorb , the decay heat release for many hours after a plant transient
- or a'cident without the need for operator action. The.
pd d?e teat sink capability of the Mark III suppression; pool l is unharized in Table 15D.2-2 and shown in Figure 15D.2-4.. Those data are based on the initial condition of.a design- { basis loss of coolant accident (DBA) except that no active supptassion pool cooling is assumed. 'Because the calculation continaes well beyond the initial blowdown resulting.from the DBA, the results are. equally applicable to a'less severe transient which results in a blowdown to the suppression t i
'15D.2-7 230-C5
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.1.3.1 Vessel Heat Removal (Continued) pool and the maintenance of low RPV pressure for a long period of time. While still conservative, the results differ from other evaluations of containment pressure provided in Subsection 6.2.1 and 15D.3, Appendix F, due to more realistic assumptions on the relation between con-tainment airspace temperature and suppression pool water temperature. As shown in Figure 15D.2-4, the initial blowdown energy (consisting of stored vessel heat, 120 seconds of rated feedwater flow, vessel sensible heat, and 200 seconds of decay heat) increases the suppression pool temperature by about 60 F and causes little change in the containment (wetwell) air space pressure (after vacuum breakers have equalized drywell and containment pressures). As decay heat continues to be discharged to the suppression pool through the safety relief valves, the suppression pool temperature increases to a point where boiling of the suppression pool may begin. At this point, slightly more than twice the initial blowdown energy has been added to the containment. i After suppression pool boiling begins, the decay heat is transferred to the containment air space and structures. As shown on Figure 15D.2-4, the containment design pressure of l 30 psia is not reached for about 27 hours. l l At this point about four times the initial blowdown energy has been transferred to the containment. The containment ultimate pressure capability extends well beyond its design pressure as discussed in Section 15D.3, Appendix G. Figure 15D.2-4 shows that this point is not reached for about 48 hours after initiation of an event in which no active heat removal from the containment occurs. Under these i l ! 230-A8 15D.2-8 L
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.1.4.1 Automatic RCIC Restart (Continued) availability of the RCIC system during transients and accidents by allowing the system to automatically restart following high vessel water level shutoff. Existing System Operation The RCIC system is described in Subsection 5.4.6. During normal plant operation the steam supply valve to the turbine is closed. Upon receipt of a vessel low water level signal, the RCIC system starts automatically. The following automatic actions occur:
- 1. The steam supply valve to the turbine opens to supply steam to the turbine. Steam line drain isolation valves then close, which isolates the RCIC steam supply from the main condenser.
- 2. Once the steam supply valve leaves the fully l
closed position the ramp generator " ramp" function is initiated. This ramp generator controls the acceleration of the turbine via the turbine control valve.
- 3. The gland seal system automatically starts.
~
- 4. Condensate suction valve remains open or is auto-matically opened to supply water to the RCIC pump.
- 5. The pump discharge valve opens to supply the water to the reactor vessel.
- 6. The cooling water supply valve opens automatically and coolant is supplied to the turbine lube oil
() cooler. 15D.2-ll 230-C6
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.1.4.1 Automatic RCIC Restart (Continued)
- 7. The test bypass valve to the cor densate storage tank closes, if initially open.
The RCIC system will automatically shut down upon receipt of any of the following signals:
- 1. Reactor high water level (see modification below)
- 2. RCIC pump low suction pressure
- 3. Turbine high exhaust pressure 4s Turbine overspeed
- 5. Auto-isolation signal
- 6. Manual turbine trip pushbutton The shutdown is affected by releasing the spring-loaded turbine trip valve. In order to reset the system it is necessary to first close the steam supply valve, then drive the motor operator of the turbine trip valve in the close direction until the spring-loaded closing latch mechanism is reset. Finally, the turbine trip valve is driven to the full open position. Closure of the steam supply valve also resets the ramp generator, closes the vessel injection valve, closes the minimum flow valve and opens the appro-priate drain valves.
Automatic Reset Modification The plannea change (Figure 15D.2-5) utilizes the steam supply valve to shut off steam to the turbine following reactor high water level, rather than using the turbine trip valve. Closure of the steam supply valve puts the system in a partial standby configuration because of the existing interlocks associated with closure of this valve. This plant change will be reflected in Subsection 5.4.6 following staff approval. 230-A12 15D.2-12
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.1.4.1 Automatic RCIC Restart (Continued) Effect of the Planned Changes The planned change will utilize the RCIC steam supply valve (E51-F045) to shut off the steam to the turbine on high vessel level rather than the turbine trip valve. The steam
, supply valve will now be used to both initiate system opera-tion at low reactor vessel water level and terminate system operation at high water level.
The time taken to shut off steam flow will be longer due to _ the nominally longer travel time of the steam supply valve compared to the trip valve. The spring-loaded turbine trip valve closes essentially instantaneously. The steam supply valve closes in fifteen seconds or less. Conservatively assuming full rated flow throughout this extended shutoff N period and a maximum rated RCIC flow of 800 gpm, an additional 200 gallons will be added to the reactor vessel following the high vessel water level trip. This volume addition has an insignificant effect on high vessel level transients including those involving high-flow rate systems (e.g., HPCS) (Subsection 15.5.1). Additional logic circuitry is added as shown in Figures 15D.2-6 and 15D.2-7. Also, an additional annunciator is added (Figure 15D.2-8) because the existing turbine trip alarm is produced by a limit switch on the turbine trip valve. The total effect on the 238 Nuclear Island design is to improve safety. The operator is no longer required to manually reset the system following a high vessel water level trip to permit later operation if needed. He will no longer be distracted by the reset action and the possibility () of inadvertent failure to reset is eliminated. The change V 15D.2-13 230-C7
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.1.4.1 Automatic RCIC Restart (Continued) utilizes the steam supply valve to terminate steam flow on high water level only. The other five RCIC trip parameters will still close the turbine trip valve requiring manual reset of the system. 15D.2.1.4.2 RCIC Break Detection Logic Modification This change is made in response to NUREG-0737 (Reference 1), Item II.K.3.15. The change increases the starting reliability of the RCIC system by reducing the likelihood of an inadvertent trip during system startup. Existing System Operation Each RCIC steam supply line is provided with two normally open isolation valves (E51-F063 and E51-F064). These valves close automatically upon receipt of an isolation signal. Each line contains a flow metering device located downstream of the isolation valves. The flow sensing system will initiate closure of the isolation valves when the flow in that line exceeds 300% of rated. A pipe rupture can produce up to ten times rated flow. The issue raised by NUREG-0737, Item II.K.3.15 (Reference 1), is that the 300% setpoint may be momentarily exceeded during the RCIC start sequences causing unnecessary trip of the RCIC system and thus less than optimum reliability. Changing the setpoint would require extensive accident analyses involving the leak detection systems as well as the RCIC system. Addition of a time delay to the break detestion circuitry directly addresses the problem and can be designed to have no impact on the currently documented accident analyses of RCIC steam supply line breaks (Subsection 15.6.4). 230-A14 15D.2-14
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 1 () 15D.2.1.4.2 RCIC Break Detection Logic Modification (Continued) l Hardware Changes The design objectives are met by replacing the existing solid state isolation logic in each break detection circuit with time delay logic. A setpoint of at least 3 seconds, but less than the 13 seconds, will be utilized. This will involve no design changes in the differential pressure measuring devices. The RCIC system has two break detection circuits. Each circuit controls one of the two isolation valves. Both circuits in the system are to be modified. A discrete time delay circuit will be incorporated for the 238 Nuclear Island. (Figure 15D.2-9) _ The timer is started when the flow rate sensed by the flow () meters exceeds the trip setpoint. This setpoint is somewhat less than the analytical limit of 300% of rated flow. This difference provides margin for instrument errors and instrument drift and ensures that actual plant performance is within the scope of the assumptions used for the accident analyses. At the end of the timer period, system isolation will only occur if the flow meters are still reading at or above the trip setpoint. A variable 3 to 10 second time delay is planned. Preoperational testing will be performed to establish the setting for an individual plant (Section 14.2). l Effects of the Planned Change ( The design objective of the RCIC isolation systen is 1.o
, limit the radiological consequences of a steam supply line rupture. The radiological consequences of such un accident are determined by the total quantity of fission procucts
() discharged to the environment (Section 15.4). Addition of a 1 15D.2-15 230-C8 l L
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.1.4.2 RCIC Break Detection Logic Modification (Continued) time delay will not result in any change in the total reactor fluid mass release when the design basis conditions are considered. This is because a 13 second valve closure delay results from the assumption in design basis radiological calculations that no offsite AC power is immediately available. The diesel-generator start and emergency bus loading sequence is assumed to require 13 seconds and precludes any movement of the isolation valves prior to this time. The modifi-cation to the isolation system would still generate an isolation signal well before emergency power is available. There is thus no impact on the design basis analysis. Furthermore, unnecessary trips of the RCXC system will be avoided resulting in attendant improved reliability. 15D.2.1.4.3 ADS Logic Modification The existing Automatic Depressurization System (ADS) actua-tion logic will be changed to respond to Item II.K.3.18 of NUREG-0737 (Reference 1). This change will automatically depressurize the reactor vessel for those events for which high pressure systems are unavailable or unable to maintain adequate water level, but do not result in a high drywell pressure trip (e.g., loss of feedwater with insufficient water delivery). Presently such an event requires manual initiation of ADS. By incorporating this change, the availability of low pressure water delivery systems is increased. Existing Logic Design The existing automatic depressurization system (ADS) logic design is shown in Figure 7.3-2. The design requires an initiation signal consisting of concurrent high drywell 230-A16 15D.2-16
- i GESSAR II 22A7007
! 238 NUCLEAR ISLAND REV. 7 15D.2.1.4.3 ADS Logic Modification (continued) i pressure and low reactor water level signals in order to actuate the ADS. The high drywell pressure signal is sealed into the initiation sequence and does not reset if the high drywell pressure subsequently clears. When both high drywell
] j pressure and low water level signals have been received, the logic confirms the water level is indeed below the scram ! water level (to prevent spurious actuations) and starts the ; 120 second delay timer. The timer is automatically reset if the low water level trip clears before the timer times out; i . it can also be manually reset. The timer allows the operator ! time to bypass the automatic blowdown if reactor water level l has been or is being restored, or if the signals are erroneous. f To complete the sequence, the ADS logic receives a low pressure ECCS permissive based on pump discharge pressure to ! provide some assurance that makeup water will be delivered l to the vessel once it is depres92rized. l An event such as a loss of feedwater may not cause a high drywell pressure signal. The ADS system is manually initiated, i if required, for such events. l Planned Logic Change Details of this logic change are not-finalized.at this time. l General Electric is currently reviewing several alternative logic changes which would accomplish the desired objectives. This. evaluation' includes an assessment of the reliability of each alternative in providing the initiation signal and avoidance of spurious initiation or other adverse effects. These results will be reviewed with the NRC, and once the final design change is selected by GE and approved by the NRC, it will be reflected in Subsection 7.3.1.1.1.2. l 15D.2-17. l 230-C9
.- . ~ . ._
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.2 BWR Mitigation Features and Capabilities This section describes the 238 Nuclear Island plant features which would mitigate the consequences of degraded core conditions in the extremely unlikely event of a severe accident. It is not expected that these features will be required to perform their functions since the systems described in Subsection 15D.2.1 are capable of preventing core damage. However, as further demonstration of the defense-in-depth approach taken in the 238 Nuclear Island design, the plant mitigation features and capabilities under postulated severe accident conditions are described below. These capabilities extend well beyond NEC requirements. One of the most significant conclusions in the 238 Nuclear Island Probabilistic Risk Assessment (Section 15D.3) is in the area of accident mitigation. Specifically, the offsite consequences of a severe accident, even one postulated to involve loss of the primary containment integrity, are decades lower than previously estimated by WASH-1400 (Reference 3). The 238 Nuclear Island containment employs a unique multi-building, multi-barrier design. The reactor vessel is enclosed in a steel and concrete drywell structure and surrounded by the pressure suppression pool. The drywell and suppression pool structures form the initial barrier around the reactor. These structures are fully enclosed in both a containment building and a shield building, which form a second and third barrier. (Figure 15D.2-10) The function of the containment system in a nuclear power plant is to protect the public from excessive dose in the event of a severe accident. In the case of the 238 Nuclear Island, this function is accomplished in two ways. The 230-A18 15D.2-18
I GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 () 15D.2.2 BWR Mitigation Features and Capabilities (Continued) first is through the multiple containment barriers; which are designed to maintain their integrity for all design basis events. Their design also provides sufficient margin to maintain their integrity for most events beyond the design basis. The second way of performing the containment function is through filtration of radioactive releases - which provides an additional level of protection for e' vents well beyond the design basis. 8 T Effective filtration, or scrubbing, of potential releases from the containment is an inherent safety feature of the I Mark III pressure suppression containment. Filtration in a 238 Nuclear Island containment is provided by both the Standby Gas Treatment System and by the suppression pool. % ,; Potential releases from the primary system, resulting from- ; ( degraded transient or accident events, pass through the suppression pool before reaching.the containment building.
~
The suppression pool effectively retains halogens and parti-culate fission products. The suppression pool retention is
~~
in addition to retention by natural plate-out mechanisms and These retention mechanisms are summarize'd containment sprays. , in Figure 15D.2-11. The 238 Nuclear Island Probabilistic - Risk Assessment has shown that the suppression pool scrubbing , u function will be maintained even in extreme accident. sequences I which might result in loss of integrity of the primary _ _ containment building. 'This arises from the substantial x3 A [ structural strength of the drywell and suppression pool. .. 7 This provides high assurance that the containment function,- 1[ 'E protecting the public from excessive dose, would be perf5rmed; even if the outer barriers (containment and shield buildings')' were to lose their integrity. d'
\ %
+
230-A19 15D.2-19
._ . - . _ _ ,- _. __ . ! 2 - - - - _ .
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.2 BWR Mitigation Features and Capabilities (Continued) The mitigative capability of the 238 Nuclear Island Contain-ment system for severe accidents can be quantified in terms of system pressure capability and fission product retention. The pressure capability of various structures within the Mark III Standard Plant Containment System are listed in Table 15D.2-3 and are described in Section 15D.3, Appendix G. As shown in Table 15D.2-3, the ultimate pressure capability of the containment extends well above its design pressure. The structures with the highest pressure capability for dynamic and static overpressurizations are the drywell and suppression pool. Therefore, whatever the pressure challenge, suppression pool and drywell integrity would likely be maintained, thereby maintaining containment function. Fission product retention is provided by suppression pool scrubbing, and other natural retention mechanisms. Quanti-fication of pool scrubbing factors is described in Subsections 15D.2.2.1.2 and 15D.2.2.3. Figure 15D.2-12 shows the results of a realistic calculation of the offsite doses for a severe accident in which the particulate fission products have been effectively retained by the suppression pool. For this calculation a full core meltdown with no system recovery was assumed. Fission products were assumed to be released at a height of 40 metters four hours after event initiation. No credit was taken for eva~Ja* ion of the population. The resulting doses are comparable to the 10CFR100 (25 rem) limit for design basis accidents. Therefore the realistic offsite doses for ] a severe accident in which all systems are assumed to fail and containment integrity is lost are comparable to doses 15D.2-20 230-C10
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.2 BWR Mitigation Features and Capabilities (Continued) conservatively calculated for design basis events where all [ safety systems continue to function with the exception of a single active component failure. Thus, the m.intenance of containment function assures no adverse offsite health effects. Figure 15D.2-12 illustrates quantitatively how the i mitigation features of the 238 Nuclear Island protect the public from excessive doses and reduce the risk from severe accidents. 15D.2.2.1 Effect of Suppression Pool Scrubbing on Severe Accident Consequences The BWR/6 probabilistic risk assessment (Section 15D.3) demonstrates that fission products would be transported to the suppression pool for nearly all accident sequences. O This section describes the effect of suppression pool scrubbing on the offsite consequences from severe accident sequences. 15D.2.2.1.1 PRA Results The distribution of events which can cantribute to the frequency of core damage are listed in Table 15D.2-4. Transient-initiated events contribute 99% of the assessed frequency of core damage of which 88% are initiated by loss of offsite power. Anticipated transients without scram (ATWS) contribute only 1.3 percent to the core damage frequency and the frequency of loss-of-coolant accidents (LOCA's) is negligible. Table 15D.2-5 lists the percent contribution to the assessed frequency of core damage by fission product release path and suppression pool condition. These results indicate that the most probable core damage event in a BWR/6 O results in fission product releases'through the safety 15D.2-21 l 230-C25
GESSAR II 22A7007 238 NUCLI:AR ISLAND REV. 4 15D.2.2.1.1 PRA Results (Continued) relief valves to a subcooled (condensing) suppression pool. These conditions would provide the highest amount of fission product scrubbing by the suppression pool. A very small fraction of the potential core damage events could result in discharge of the fission products into a saturated (non-condensing) suppression pool. Since a thermally saturated suppression pool represents a worst case condition for fission product scrubbing, GE's scrubbing tests were performed to simulate this condition. Even under these limiting conditions, the suppression pool was found to provid,.e an extremely high fission product retention capability as discussed in Subsection 15D.2.2.3. The next subsection summarizes the results of a survey of the available literature on suppression pool fission product scrubbing. The results of recent experiments performed by GE are provided in Subsection 15D.2.2.3 and Attachment A. 15D.2.2.1.2 Literature Survey of Suppression Pool Scrubbing Factors 15D.2.2.1.2.1 Introduction In severe accident sequences, the presence of water in the fission product transport pathways provides an important means to minimize the quantity of airborne fission products. The 238 Nuclear Island uses the pressure suppression pool to provide a water barrier to fission product migration. Thus, significant retention of radiciodines and other fission products, except noble cases, is expected and must be accounted for in any realistic evaluation of accident consequences. O 230-A22 15D.2-22
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 O 15D.2.2.1.2.2 Summary of Results of Literature Survey (Continued) which the existing data base could support, and the potentially attainable DFs which could be supported by further testing, were presented for each dominant i transport sequence in Reference 6. 15D.2.2.1.2.3 Conclusions from Literature Survey Results of the literature survey (Reference 6) indicate that chemical forms similar to the inorgaaic iodides and particu-lates that would be expected to be released during postulated severe accidents would be retained in the suppression pool and would not escape into the primary containment air space. Suppression pool decontamination factors that were found to j O be appropriate for use in BWR- risk assessments, based on the literature survey, are presented in Table 15D.2-7 (repeated from Reference 6). Based on the data presented in NEDO-25420 and the expected BWR transport conditions, it was concluded that suppression pool decontamination factors of at least ,
- 102 for elemental iodine and 103 for particulate and cesium l iodide could be expected for subcooled pools. For saturated pools, decontamination factors of at least 30 for elemental iodine and 102 for particulates and-cesium iodide were l justified by the literature data. NEDO-25420 also concluded that minimum values in Table 15D.2-7 would be increased several orders of magnitude by testing for the range of expected conditions during severe accidents. Testing has recently been completed by GE and the results, which are described in Attachment A, demonstrate that the suppression pool DFs arc orders of magnitude higher than the lower bounds established by the literature survey.
15D.2-27 230-C11
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.2.2.1.2.3 Conclusions from Literature Survey (Continued) Other natural processes such as the agglomeration of parti-culates, plateout, deposition and washout also play an important role in limiting the quantity of fission products available for leakage to the environment. The overall attenuation factor applicable to BWR severe accident sequences includes both the effects of pool scrubbing and natural removal processes expected to occur in the various compartments of the 238 Nuclear Island. 15D.2.2.1.3 Effect of Pool Scrubbing Factors On Offsite Doses A recent NRC review (Reference 5) of fission product releases in severe accidents concluded that all fission products other than noble gases and methyl iodide would be released as aerosols (i.e. particulates suspended in a gas phase of steam and noncondensables). As concluded in Subsection 15D.2.2.1.1, postulated severe accidents leading to core damage in the 238 Nuclear Island result in the transport of fission products to the suppression pool. Fission product aerosols discharged to the pool must pass through 13 to 19 feet of water in the pool before reaching the primary containment airspace. Additionally, any aerosols escaping the pool would have to rise approximately 130 feet vertically to reach the most probable containment release path. Sequences which would allow all fission products to bypass the pool require failures in addition to ; the multiple failures which lead to core damage and are so I highly unlikely as to be incredible. These bypass scenarios are discussed in Section 15D.3. O 230-A28 15D.2-28
GESSAR II 22A7007 230 NUCLEAR ISLAND REV. 7 15D.2.2.2.1.1 Scrubbing Mechanisms (Ccntinued) (3) Brownian Diffusion (Continued)
~
kd = 1.8 Df 1/2 VR3 where: - k d
= diffusive absorption coefficient, cm -1 D = diffusivity due to Brownian motion f
(Df = kTC, ), cm2/sec 3np d g R = bubble radius, cm k = Boltzmann constant, 1.38 x 10 ~10 g-cm 2 /sec t oK T = temperature in *K The mobility of the particles decreases with increasing-particle size, and the diffusive abs 9rption coefficient
- Os is generally negligible compared to inertial and sedi-mentation adsorption coefficients for a particle size
>1 pm.
l 15D.2.2.2.1.2 Overall Particle Absorption Coefficient and l Decontamination Factor The total theoretical particle absorption coefficient for the particle in a gas bubble scrubbing process is Ki=kni + k,g + kdi l and the rate of particle absorption per unit path of the bubble in a water column is given by: 1 dC = -K C l 3 3$ r T 15D.2-39 230-C12
-n., _, , _ , - - - - -
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.2.2.1.2 Overall Particle Absorption Coefficient and Decontamination Factor (Continued) Upon integration,
~
C. = C? e
-Kf L 1 1 .
where ci = particle concentration of species i in gas bubbles at the outlet of a water column, o Cy = particle concentration of species i in gas bubbles at the inlet of a water column, 1 = height of water column, cm The decontamination factor (DF) for the scrubbing process is the ratio of the inlet concentration to the outlet concentration U C Kly
- 1 * *
(DF)1 Cg For a mixture of particles with various particle sizes (but same particle der.sity), the overall mass or activity DF can be calculated by: (DF) overall
- Ig(Fg DF)i) where Fg = the mass or activity fraction o trticle size, i, in the mixture at gas t ,le inlet 15D.2-40 O
230-Cl3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.2.3 Suppression Pool Scrubbing Factors for Severe Accidents 15D.2.2.3.0 Introduction Using the combined particulate scrubbing (Subsection 15D.2.2.2.1) and hydrodynamic model (Subsection 15D.2.2.2.2), it is possible to develop a method for calculating the decontamination factor expected in the suppression pool during severe accidents. A discussion of this methodology is presented here. Hydrodynamic theory (Subsection 15D.2.2.2.2 and Attachment A) and tests demonstrated that the bubbles rise through the suppression pool in a swarm of small bubbles. The decontamination factor for the bubble swarm is equal to O the summation of the fractional contributions of all the bubbles. The decontamination factor for each bubble is governed by its scrubbing height. The simple calculational model which sums the decontamination factor contributions of the small bubbles in a swarm for a particle size, is shown in Figure 15D.2-14. , Three calculation cases are presented in Subsection 15D.2.2.3.2. The calculational procedure uses the equation given in Figure 15D.2-14 (where the terms have been previously defined in Subsection 15D.2.2.2.1) to calculate the decontamination factor as a function of particle size in the bubble swarm. i The total decontamination factor (DF ) is determined by T l summing the particle size DFs over the particle size distribution. 1 () T I g(M /DFi) ( f _. 15D.2-45 230-C14
GESSA2 II 22A7007 ; 238 NUCLEAR ISLAND REV. 4 l 15D.2.2.3.1 Model Inputs Particle size distributions for corium-steel and corium-concrete experiments were obtained from Sandia Laboratories I (References 12, 13) and were used for the calculations in l Subsection 15D.2.2.3.2. They are shown in Figures 15D.2-15 ) and -16. Bubble volumes were calculated using the accident sequence dependent flow rates of steam and non-condensibles
)
obtained from the MARCH Code (Section 15D.3, Appendix F). 15D.2.2.3.2 Calculated Scrubbing Factors For 238 Nuclear Island Total decontamination factors using the particle size distributions in Figures 15D.2-15 and 16 and considering dic;harge into the suppression pool from x-guenchers and < horizontal vents were calculated. The calculated results are shown in Table 15D.2-15. The model calculation for the fission product scrubbing conditions is conservative (smaller DF than would be expected under actual accident conditions). Under the postulated accident conditions, steam and hydrogen would dominate the gas phase in the bubbles. The gas properties of a steam / hydrogen mixture were not taken into account in these model calculations. The absorption efficiencies due to particle inertia, sedimentation and diffusion should increase for steam / hydrogen compared to air. Furthermore, the effect of steam condensation is not included in the model. In a most recent study (Reference 14), the DF has been predicted to increase nearly linearly or exponentially, depending on the particle size, with increasing steam content in the gas bubbles. O 230-A46 15D.2-46
._ - - - - - - - - - . . = .
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 () 1SD.2.2.3.3 Conclusions Regarding Pool Scrubbing Factors Using particle distributions from corium-steel and corium-concrete meltdown experime.ts, the estimated scrubbing factors are DF>104 for discharge through the quenchers and DFS10 2 _lo4 for vent discharges. These results confirm that the BWR suppression pool would effectively retain fission ] product particles released under severe accident conditions. i
- 15D.2.2.4 Conclusions Subsection 15D.2.2 provides a description of the 238 Nuclear Island features and capabilities which mitigate the consequences of postulated severe accidents. These capabilities extend well beyond NRC requirements.
The 238 Nuclear Island pressure suppression pool filters
/ potential fission product releases, thereby limiting offsite
)
4 doses and maintaining containment function. The BWR/6 Standard Plant Probabilistic Risk Assessment has
- shown that suppression pool scrubbing will be maintained even in extreme accident sequences which might result in loss of integrity of the primary containment building. This results because of the substantial strength of the drywell j and suppression pool structures.
Quantification of the fission product scrubbing capabi-lity of the suppression pool during-severe accidents was i accomplished by GE's Fission Product Scrubbing Program. This program resulted in the development of a first principles [ analytical model to describe pool scrubbing and the experimental , verification of the model by_ mass-transfer and hydrodynamic testing. This model predicts that the suppression pool i 1SD.2-47 4 230-C15
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.2.4 Conclusions (Continued) would reduce particulate fission product releases by a factor of 10,000 in the unlikely event of a severe accident. These results confirm that the 238 Nuclear Island suppres-sion pool would effectively retain fission products releases during severe accidents. This backup mitigative capability to maintenance of containment integrity provides further defense in-depth for assuring that the health and safety of the public is protected. 15D.2.3 Emergency Procedure Guidelines one of the tasks undertaken by General Electric and the Boiling Water Reactor Owner's Group in response to post-TMI NRC requirements was to develop generic emergency procedure _ guidelines (EPGs) (Reference 2). The first version of the EPGs (Revision 0) was developed for BWR/1-5 inventory threatening events and submitted to the NRC in June 1980. The EPGs were extended to cover BWR/6 in Revision 1 which was submitted to the NRC in January 1981. Revision 1 of the EPGs is applicable to the 238 Nuclear Island. Revision 2 in prepublication form was provided to
~
the NRC on June 1, 1982. Revision 2 is also applicable to the 238 Nuclear Island. , I The EPGs are symptom-based guidelines as opposed to event-based guidelines. The operator does not need to identify what event is occurring in the plant in order to decide on what actions to take. Rather, he observ'se the symptoms (utilizing a relatively few instruments) which are , occurring and takes immediate actions based on responding to i
- l The symptom based EPGs provide a major those symptoms.
15D.2-48 230-C16 l l
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7
/~'\ 15D.2.3 Emergency Procedure Guidelines (continued)
O simplification over previously developed procedures because for inventory threatening events, the operator needs only to maintain reactor vessel water level, independent of the reason for an initial water level decrease. These symptom-based guidelines are a significant improvement since all of the existing emergency procedures used at BWRs, prior to development of the EPGs, were event dependent. If the operator did not have an emergency procedure for the i event which was occurring at his plant, he may not have known the correct action to take. Operator confusion was possible since there were only a limited number of emergency procedures in place, but a large number of possible events. Symptomatic EPGs avoid that problem. There are a limited . number of important symptoms and proper response is outlined () in the EPGs. The specified operator actions depend upon the
\ ability the operator has to monitor and respond to each
]
symptom or instrument reading.
- The other significant improvement to existing emergency ,
procedures is that the EPGs have been developed with , extensive GE - owner interaction. Realistic analysis has been used repeatedly when making decisions as to what steps are appropriate for inclusion in the EPGs. This analysis is used to decide on various action, levels for the symptoms l and in identification of the success paths for total response to these symptoms. . The EPGs also go beyond the design basis of plant licensing. They are not limited to defined dcuign basis events or single failures. The EPGs cover the range from all equipment (including non-safety grade systems) available to every
/~'T injection system unavailable.
O 15D.2-49 230-C17
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.3 Emergency Procedure Guidelines (continued) A summary description of the content of NEDO-24934 (EPG Revision 1) follows. EPG Revision 1 contains guidelines on level control, shutdown, and containment control. The Level Control Guideline section provides guidance for , operator use in restoring and stabilizing reactor pressure . vessel (RPV) water level. The plant symptoms which alert the operator to enter this guideline are low RPV water level, high drywell pressure, or an isolation has occurred. The shutdown section provides guidance for operator use in depressurizing the RPV to cold shutdown conditions. This guideline is entered from the Level Control Guideline after the RPV water level has been stabilized. The Containment Control section provides guidance for the
~
operator to use in controlling primary containment tempera-tures, pressure, and level whenever suppression pool temper-
~
ature, drywell temperature, containment temperature, drywell pressure, or suppression pool water level are above their _ normal operating limit or suppression pool water level is below its normal operating limit. This guideline is executed concurrently with the guideline from which it is entered. Cautions are noted at various points throughout the guide-lines. These cautions clearly remind the operator of important instructions. l The EPG also contains contingencies for use when the symptoms are not being satisfactorily controlled as a result of actions taken from the main guidelines or as a result of
}
l 15D.2-50 i 230-C18
GESSAR II 22A7007 238 NUCLEAR ISLAND REY. 7
- 15D.2.3 Emergency Procedure Guidelines (continued) iv equirment malfunctions. The contingencies included are Level Restoration, Rapid RPV Depressurization, Core Cooling without Injection, Core Cooling without Level Restoration, Alternate Shutdown Cooling, and RPV Flooding.
In summary, the EPGs are symptom-based guidelines which simplify operator response. They enable the operator to j respond to symptoms rather than requiring that he diagnose the cause of the event. Their applicability extends beyond design basis events, and their development represents a significant improvement over past emergency procedures. The EPGs, in addition to the inherent 238 Nuclear Island features, provide an additional level of BWR operational safety relating to the prevention of severe accidents. 15D.2.4 References ,O
- 1. U. S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," USNRC Report NUREG-0737,
( November 1980.
- 2. NEDO-24934, " Emergency Procedure Guidelines .BWR 1-6,"
General Electric Company, January 1981.
- 3. " Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400-(October 1975).
- 4. Regulatory Guide 1.3, " Assumptions Used for' Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors," June, 1974.
230-A51 15D.2-51
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.4 References (Continued)
- 5. NUREG-0772, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents," USNRC, June 1981.
- 6. Rastler, D. M., " Suppression Pool Scrubbing Factors for Postulated BWR Accident Conditions", NEDO-25420, Class I, General Electric Company, June 1981.
- 7. Stratton, W. R. et.al, Letter to NRC Chairman John Ahearne on August 14, 1980.
- 8. Cohen, B. L. and Lee, I. S., "A Catalogue of Risks,"
Health Physics, 36 (1970) pp. 707-722.
- 9. Grosch, D. S. and Hopwood, L. E., " Biological Effects of Radiation", Academic Press, Inc. (1979).
- 10. N. A. Fuchs, "The Mechanics of Aerosols," translated from the Russian by R. E. Daisley and Morina Fuchs, Pergamon Press (1964).
- 11. S. Yuu, T. Jotake, and K. Abe, Power Technology, 17, 115-122 (1977).
- 12. NUREG/CR-0324/ SAND 78-1511 " Light Water Reactor Safety Research Program, Quarterly Report, January-March, 1978," P. 17-24.
- 13. NUREG-0183-5/ SAND 78 - 0076, " Light Water Reactor Safety Research Program, Quarterly Report, July - September, 1977, P. 8-12.
O 230-?.s2 153.2-52
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.2.4 References (Continued)
- 14. D. C. Bugby, A. F. Mills, and R. L. Ritzman, " Fission Product Retention in Pressure Suppression Pools,"
Science Application, Inc., January 7, 1982.
- 15. BWR Emergency Procedure Guidelines (prepublication form)'
submitted in Letter BWROG 8219 dated 6/1/82 from T. J. Dente (BWR Owners Group) to D. G. Eisenhut (NRC). l O l l O- 15D.2-52a 230-C21
. - - - - .= .
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 TABLE 150.2-1 BWR/6 WATER INJECTION SYSTEMS CAPABILITY Total Flow Per Pump - High Pressure No. of Pumps (apa) Necessary Conditions ! HPCS* 1 1550 3 RCIC 1 800 FW 2 17000 Offsite Power & Condensate, CRD 2 85 Offsite Power i SLC 2 43 . Low Pressure HPCS* 1 6000 ) RCIC 1 800 RPV Pressure > 50 psi . FW 2 15000 Offsite Power CRD 2 85 Offsite Power . SLC 2 43 "' LPCS* 1 6000
- LPCI* 3 7100 FW Cor.densate 3 11500 Offsite Power l RHR Service 2 300 _
Water l
- Emergency Core Cooling Systems (ECCS) l
( l l 150.2-53 229Al
i GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 I TABLE 15D.2 MARK III LONG-TERM PRESSURIZATION l l Total Suppression Btu Added Pool Water Suppression Containment Time to Suppression Mass Pool Temp Pressure After LOCA Pool 108 Btu 106 lbm *F psia sec (hr)
- 0. O. 8.06 80. 14.7 200. 847.* 11.29 141. 14.7 (0.06) 1731. 952. 11.29 150. 15.0 (0.48) 12294. 1361. 11.28 185. 15.6 (3.4) 25694. 1756. 11.26 218. 16.3 (7.1) 98544. 3349. 11.08 250.** 30.
(27.4) 171244. 4591. 10.84 305.** 72. (47.6)
- Includes: Upper Pool Energy = 196 x 106 All Vessel Inventory = 334 x 106 120 sec. of Rated Feedwater = 206 x 106 Vessel Sensible Heat = 85 x 106 200 sec. of Decay Heat = 26 x 106
** Assumed at Saturation temperature for containment pressure.
O 150.2-54 229J2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 O HIGH PRESSURE l 4 FEEDWATER J
-c - CORE SPRAY C ISOLATION COOLING 13 PUMPS 4 CRO COOLING LOW PRESSURE
^ 4 COOLANT INJECTION AND FLOOD
= CORE SPRAY
! C CONDENSATE ACTUAL BWR CAPABILITY l l Figure 15D.2-1. Systems to Supply Water to
- Core 1
150.2-71 L -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 O STEAM SAFETY / RELIEF VALVE I g > MAIN STEAM WATER LEVEL N SUPPRESSION POOL NATURAL CIRCULATION + STEAM RELIEF
= PASSIVE DECAY HEAT REMOVAL Figure 15D.2-2 Decay Heat Removal 150.2-72
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 O I l l l J_
/.
4.
,1 = AP w
/ ,
CORE i f/' , ~
.11, .Ly e DIRECTLY ON VESSEL
/
- AP MEASUREMENT I
- REDUNDANT l
l o nw e isD.2-3. w e er Levet ne - e m e 15D.2-73
N La sooo - m EXPECTED 4591 72 @ LOSS OF 4g ag m j g CONTAINMENT () tij P INTEGRITY M CO Y ?*
~
==
= 3349 30 pda CONTAINMENT HH J DESIGN - 27 U2 H O PRESSURE O 2 3000 - 6 5 O b Z r z o o DECAY $ W
$ HEAT O $ $M p
1756 2180F SOILING 16.3 psie 7 I m STARTS m N E 0 l F 1000 - 847 1410F 14,7 p.g. g _ 8 LOWDOWN o o M NN HEAT ADDITION SUPPRESSION POOL PRIMARY (p p TO CONTAINMENT TEMPER ATURE (OF) CONTAINMENT < wi PRESSURE
- O (psie) y Figure 13D.2-4. Containment Passive Heat Sink Capability 9 9 O
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9
, , , . i i s ui i = i = ia i -.
i . i i p, . i Figure 15D.2-8. Planned RCIC , Restart Modification (Figure 7.4-la) 15 D . 2-78 a/15D .*2-70b
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 N [ . 5 i 0 ' f 1 , lL_,_ '. . U
. l '
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- c1, ,mn ~*, .?in ':W:,
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REGULATORY ACTUAL 8 ASIS CAPABILITY DECONTAMINATION FACTORS
- PLA TEOUT 2 10
- SUPPRESSION POOL 1 100-10000
- CONTAINMENT SPR AYS 2 10 10000
- SECONDARY CONTAINMENT 5 5
- STANDBY GAS TREATMENT 100 1000 SYSTEM TOTAL 2000 50,000,000 Figure 15D.2-ll. Mark III Fission Product Retention (Halogens and Particulates) 15D.2-81
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 O EVALUATION BASIS WAI I e CORE MELT SOURCE TERMS 10.000 - op.i e CONTAINMENT FAILURE AT 4 HOURS e NO EVACUATION 1,000 - FATALITY THRESHOLD (320 rem) 2 su 100 - 10CFR100 LIMIT (25 rem)
,a _
REALISTIC EVALUATION i i i i l I 1 l l 1 g 2 3 4 5 6 7 8 9 0 1 MILES DOWNWIND FROM SITE l Figure 15D.2-12. Importance of Fission Product Retention in Mark III Pressure Suppression Containment 15D.2-82
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.4.2.1 Comparison of-PRA Results to Draft Safety Goal (} (Continued) Using the PRA results in Section 15D.3, comparison to the NRC proposed guidelines is provided in Table 15D.4-1. Comparison is made to all the numerical guidelines dealing with mortality risks and plant performance.
-6 The calculated core melt probability of $5 x 10 for the 238 Nuclear Island is a factor of 20 below the proposed guideline. As noted in Section 15D.3, there were no cal-culated early (prompt) fatalities, consequently, the 238 Nuclear Island design results are well below the NRC guide-lines for individual and societal prompt fatality risks.
The NRC numerical guideline for individual latent fatality risk is based on 0.1% of national statistics and is equiva-
-6 lent to s2.0 x 10 . The 238 Nuclear Island value is more than four orders of magnitude below this guideline. The PRA
[} result of 2 x 10
-4 latent fatalities (the mean of the risk curve for fatalities within 500 miles) when divided by the population within 50 miles of the site (s8.2 million people)
-11 yields the individual latent fatality risk of 2.5 x 10 ,
The societal latent fatality risk from Section 15D.3,
-4 latent fatalities, is a factor of 10 ~4 below the 2 x 10 guideline value of 3.2.
These results illustrate the capabilty of the 238 Nuclear Island design, as quantified in Section 15D.3, to meet the proposed NRC guidelines with extensive margin. v
')
230-B7 15D.4-7 l
w GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.4.2.2 Consideration of Hydrogen Control and Containment Design Changes one of the key issues identified b/ the NRC related to ^ severe accidents has been hydtcren control. In the notice of Proposed Rulemaking on Inte2 1 Requirements Related to ~ Hydrogen Control in December, l81, the NRC has proposed that additional hydrogen cortrol systems be added to the , BWR/6 - Mark III design to accommodate hydrogen release from ?' postulated degraded core accidents. The proposed rule would require applicants to demonstrate maintenance of containment integrity for events which release an amount of hydrogen equivalent to 75% metal-water reaction of the active fuel cladding. GE has provided detailed responses to this proposed rule (Reference 3 ) . -
]
The Probabilistic Risk Assessment in Section 15D.3 assessed hydrogen generation for severe accidents. Further, the PRA quantified the consequences of hydrogen combustion events ' taking account of the structural capability of the drywell and pool to assure pool scrubbing of potential releases even if containment integrity were lost. Thic provision of fission product retention via suppression pool scrubbing means that containment function is maintained even for severe accidents. GE believes that maintenance of containment function should be the acceptance criteria for hydrogen control following severe accidents. Only a minimal risk reduction could be realized by eliminating hydrogen combustion accident sequences. The net effect of precluding hydrogen combustion for certain degraded core accident sequences would be to shift the loss of containment inte-grity from the time of hydrogen combustion to the time of containment overpressurization from non-condensibles generated by the core-concrete interaction. i 15D.4-8 l 230-C22
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15D.4.2.2 Consideration of Hydrogen Control and Containment Design Changes (Continued) This delay in the time of fission product release reduces the 238 Nuclear Island risk by less than 30%. This small reduction is a result of additional time for fission product decay. Therefore, in relative terms, the addition of a hydrogen control system provides only a minimal risk reduction. On an absolute basis, the 238 Nuclear Island risk is already low compared to the proposed NRC Safety Goal (Subsection 15D.4.2.1), and thus the provision of an additional hydrogen control system is inappropriate. In addition to hydrogen control, the NRC has considered other changes affecting the Mark III containment design for ) future plants. Specifically, the NRC has considered the s/ appropriateness of increasing the Service Level C capability for all containment designs to 45 psig. The NRC has also considered the appropriateness of. including one or more dedicated containment penetrations in order not to preclude < future installation of systems to prevent breach of contain-ment, such as a filtered vented containment system. The following paragraphs provide a discussion of the significance of these proposed changes relative to the existing capability of the 238 Nuclear Island. The 238 Nuclear Island mitigation features were presented in Subsection 15D.2.2 and the containment design capability was presented in Section 15D.3 Appendix G. These containment evaluations show that, in addition to the ultimate pressure capability of the primary containment significantly exceeding 45 psig, the Service Level C capability for the drywell and (s
\
pool structures also significantly exceeds 45 psig. on the drywell and suppression pool strength, it has been Based 15D.4-9 230-C23
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15D.4.2.2 Consideration of Hydrogen Control and Containment Design Changes (Continued) concluded that the fission product retention function of the containment will be maintained. Maintenance of this function is accomplishad even for postulated severe accidents with loss of integrity of the primary containment. Thus, fission product retention will be assured as a result of the structural capability of the drywell and pool. A further increase of the prmary containment building structural capability would not significantly reduce the risk due to severe accidents, since the non-condensible gases generated from degraded cores may still ultimately pressurize the containment above any increased capability value. With respect to assuring that penetrations for a filtered vent could be provided, it has been shown in Subsection 15D.2.2 that the pressure suppression pool effectively filters fission product releases from postulated severe accidents. The pool and drywell thus provide the same mitigation capability as a filtered vented containment system, even for U ents that proposed filtered vented containment system lesigns cannot accommodate. Therefore, it is concluded that changes to the pressure capability and the provision of a dedicated penetration are inappropriate for the 238 Nuclear Island containment. 15D.4.2.3 Consideration of Proposed NRC Policy Statement (SECY-82-1) In the proposed policy statement included as Attachment A to SECY-82-1 (Reference 2), implementation guidelines are offered for new Construction Permit (CP) applicants. Although these guidelines are only in draft form, they provide useful criteria for measuring the capabilities of the 238 Nuclear 230-B10 15D.4-10
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 TABLE 15D.4-1 O COMPARISON OF 238 NUCLEAR ISLAND PRA RESULTS TO PROPOSED NRC SAFETY GOALS Proposed NRC 238 Nuclear Island Criteria Guideline Result
~4 +5.0 x 10
-6 Core Melt 1.0 x 10 Probability Individual Prompt 5.0 x 10 ~7 III 0(5)
Fatality Risk Individual Latent 2.0 x 10 -6 (1) 2.5 x 10 -11 (4) Fatality Risk O Societal Prompt 1 x 10 -4 (2) 0(5) Fatality Risk
~4 Societal Latent 3.2 (3) 2 x 10 Fatality Risk NOTES:
(1) 0.1% of National Fatality Statistics ] (2) Assuming 1 mile average population of l' Jeople. (3) Assuming 50 mile average population of 1. million people
~4 deaths spread over the PRA (4) Using theoretical 2 x 10 site 6 - 50 mile population of 8.2 million people (5) No prompt fatalities were calculated for the 238 Nuclear Island PRA 15D.4-15/15D.4-16 230-C24
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4
- TABLE 15DA.1-1 PARTICLE SIZE DISTRIBUTIO.1 ON IMPACTOR STAGE MAIN STREAM IMPACTOR (INLET)
PLATE NO. OF CUNNINGHAM PARTICLE NO. HOLES HOLE DIAM, CM SLIP DIAM, CM 0 264 0.16130 1.05103 3.196E-04 1 264 0.11810 1.08266 1.973E-04 2 264 0.09140 1.12369 1.318E-04 3 264 0.07110 1.18515 8.807E-05 4 264 0.05330 1.29861 5.461E-05 1 5 264 0.03430 1.65624 2.496E-05 6 264 0.02540 2.22490 1.372E-05 7 156 0.02540 2.87085 9.288E-06 Particle Density = 7.800 gm/cc Flow Rate = 472.000 cc/sec Temperature = 293.00 deg K TABLE 15DA.1-2 PARTICLE SIZE DISTRIBUTION ON IMPACTOR STAGE AEROSOL IMPACTOR (EXIT) PLATE NO. OF CUNNINGHAM PARTICLE < NO. HOLES HOLE DIAM, CM SLIP DIAM, CM 0 264 0.16130 1.07759 2.285E-04 1 264 0.11810 1.12663 1.400E-04 2 264 0.09140 1.19126 9.272E-05 3 264 0.07110 1.29010 6.113E-05 4 264 0.05330 1.47902 3.706E-05 5 264 0.03430 2.14666 1.588E-05 6 264 0.02540 3.40894 8.029E-06 7 156 0.02540 4.95946 5.117E-06 Particle Density = 7.800 gm/cc Flow Rate = 944.000 cc/sec Temperature = 315.00 deg K 15DA-13 229J26
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 TABLE 15DA.1-3 COMPARISON OF TEST MATRIX AND SEVERE ACCIDENT CONDITIONS RANGE EXPECTED FOR PARAMETER RANGE TESTED SEVERE ACCIDENTS Bubble Size (cm): 0.4 - 1.4 0.5 - 0.6 Particle Concentration (g/m3 ): 0.02 - 5.5 >5.5 Submergence Height (cm): 34 - 167 411 - 573 Gas and Water Temperature (*C): 20 and 60 49 - 649 Particle Size Distribution (pm): 0.05 to 10 .05 - 10 ] O 15DA-14 229A2
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 j TABLE 15DA.1-4
SUMMARY
OF SCRUBBING TEST RESULTS Test Bubble Gr!fice/ Bubble Rate Particle 3 Weight Overall Date Diam,cm Cap (1)_ B/ Min Conc (a/m ) Height cm D. F. EU 0 - Millip re 23 9/16 0.47 70/none 183 0.18 34.3 108 9/29 0.63 70/0.6 145 0.17 34.3 333 9/30 0.63 70/0.6 145 0.44 34.3 214 10/1 0.60 70/0.6 277 0.02 34.3 119 10/8 0.74 130/0.8 254 0.53 34.3 189 10/27 0.85 180/0.8 312 0.48 167.7 1170 10/28 0.85 180/0.8 318 0.91 167.7 1415 10/30 0.45 180/0.2 248 0.87 167.7 1251 11/3 0.45 180/0,2 48 0.48 167.7 719 EU 0 - Impactors 23 12/1 0.86 180/0.7 260 0.76 167.7 896 12/2 0.86 180/0.7 260 5.5 167.7 1260 12/8 1.41 Special 124 0.30 167.7 534 12/9 1.35 Special 140 1.8 167.7 1260 12/10 0.78 180/0.7 248 4.95 76.2 910 0 12/11 12/14 12/15 0.88 0.88* 0.88 240/0.7 240/0.7 240/0.7 276 272 304 4.34 1.38
**N.D.
167.7 167.7 167.7 4157 2270 928 60*C Water /60*C Gas
** Not Determined ,
(1) Orifice size in microns, cap size in centimeters l 15DA-15 229J28 l . - - . . - - - ,_
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 TABLE 15DA.1-5 EUROPIUM OXIDE PARTICLE SIZE DISTRIBUTION Average Particle Size
- Mass Fraction (p)
.001 .1
.009 .15
.04 .33
.05 .67
.1 1.13
.1 1.73
.1 2.46
.6 > 2.46 ]
O ^ p = 7.8 gm/cma 15DA-16 0 229A3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 O 15DA.3 HYDRODYNAMIC THEORY Ihis section describes the hydrodynamic theory which explains the observations in the hydrodynamic experiments. 15DA.3.1 Bubble Breakup Mechanisms The mechanisms which contribute to bubble breakup identified in this analysis involve inviscid flattening, aerodynamic shredding, Taylor instability, and Helmholtz instability. Inviscid flattening refers to the bubble initial distortion after it is free from its charging source and before a significant wake develops below it. Aerodynamic shredding involves a pressure reduction on the bubble equator due to higher liquid velocity as the bubble moves upward, and subsequent tearing apart as surface tension forces are overcome. Taylor instability refers to the amplitude growth {/}
'- of interface waves as a gas bubble top surface supports ,
liquid above in a gravity field. Helmholtz instability involves amplitude growth of the interface when fluids of different velocity and density flow in parallel streams. Inviscid flattening of a bubble is shown sequentially in Figure 15DA.3-1. The bubble is free from its point of charging and its pressure is the average of its surrounding liquid. The bottom surface penetrates upward toward the top, tending to flatten the bubble. This phenomenon is crudely explained by the fact that the bubble pressure is uniform throughout the contained gas, whereas the total pressure of the surrounding liquid is higher at the bottom due to the hydrostatic head. However, since the bottom liquid also must be at the bubble pressure, its velocity must be higher than that of the top liquid. l O V 15DA-33 49-C47
I GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15DA.3.1 Bubble Breakup Mechanisms (Continued) ; i It is seen in Figure 15DA.3-1 that the bottom surface almost ! catches the top when it has risen about one initial radius. The rise velocity of the top surface is about 0.5/gR, and - that of the lower surface is about 1.38/gR. For a bubble of . 1.0 ft. radius, the upper and lower surfaces rise at about 2.8 and 7.8 ft/s. When a free bubble begins to rise, vortex generation occurs from viscous shear in the surrounding liquid, and a wake is created which ultimately limits the rise velocity. Figure 15DA.3-1 is based on inviscid theory and applies prior to wake formation. It is expected that the initial lenticular distortion occurs before a significant wake effect can occur. Consider a somewhat flattened bubble rising at steady velocity V, through stationary liquid as shown in Figure 15DA.3-2. An observer on the bubble would see flow coming toward him at velocity V,. A stagnation point would occur at the top of the bubble, and velocity of fluid at the outermost edge would be V. Potential flow theory shows that if the rising object were a cylinder, the ratio V/V, is 2.0. The flattened bubble of Figure 15DA.3-2 resembles the cross section of an ellipse on the top half with major and minor axes 2a and 2b, for which V/V, = 1 + a/b (1) Bubble internal pressure is approximately equal to the stagnation pressure P g , and average fluid pressure P on any quadrant roughly corresponds to the average velocity on a cylinder, namely V/n, for which 15DA-34 49-B1
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15DA.3.2 Bubble Breakup Distance (Continued) than inviscid flattening alone. Bubble rise height with all breakup mechanisms active can be estimated from energy methods. Suppose that a bubble of radius R g and volume "Vg " is initially submerged below the pool surface a distance H, as shown in Figure 15DA.3-3. This configuration corresponds to a gas-liquid system with a given value of initial energy. As the bubble rises, the system energy is redistributed between fluid kinetic and potential energies, the energy associated with surface tension as new bubbles are formed, and liquid internal energy increase due to s'scous drag effects. If gas compressibility is neglected, the liquid potential energy associated with a spherical Subble submerged to depth H as shown in Figure 15DA.3-3 corresponds to the work of l submergence, Eg=yVHg (11) If the initial bubble has divided into n equal bubbles of radius r by the time it rises to elevation y , the sum of volumes is equal to the, initial volume, which leads to n = (Rg /r)3 (12) l The liquid potential energy for n bubbles at elevation y is given by PE = n (4nr 3 /3) y (H-y) = V o y (H-y) (13) ] the liquid bulk kinetic energy of a single bubble of radius r
- moving at velocity V is (2nr 3 /3)y V 2 /2g. Therefore, the l kinetic energy of n bubbles is 15DA-39 l 49-B2 l
L
I GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15DA.3.2 Bubble Breakup Distance (Continued) O KE = n (2nr3 /3) y V 2/2g = (2nRg8 /3) V2/2g (14) 1 The initial bubble surface area is 4nR g2 The increased area when n bubbles of radius r have formed is n(4nr2 ) l l so that the surface tension energy stored in the newly l created surface area is Eg = a (n4nr 2 - 4nRg 2 ) = 4nR g 2 o(Rg /r - 1) (15) The increase of dissipation energy forms associated with vorticity and internal energy in the liquid is equal to that energy transferred as the rising bubbles perform viscous or drag work. The drag force of one bubble of radius r is ] given by C ynr V /2g. Therefore, as the number of bubbles d increases, the dissipated energy is Y O Y 1 E = (nC Ynra y2/29)dy = C YnRg V2/2g g - dy (16) g d d Since the total system energy must remain constant, Eg = PE + KE + E g +E t (17) Assuming a constant average bubble velocity dy/dt = V, the derivative of E g is written as dEg /dt = 0 = -Vg yV + 0 - (4noRg 3/r a )dr/dt + Cd ynRg 3V3/2gr 15DA-40 O 49-B3
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 7 15DA.3.2 Bubble Breakup Distance (Continued) O or, dr/dt - (C yV3/8ag)r = -(yV/3a)r2 d (18) with the initial condition, t=0, r=R g (19) and bubble elevation is given by y = Vt (20) A solution of Equation (18) combined with (19) and (20) yields r = (3Cd V /8gR,) (21) lig 1 + (3Cd V2/8gR, - 1)exp(-(CdT #0"9)Y) Equation (21) gives the size of bubbles formed at elevation y . As y increases, the bubble size becomes r + 3C v /8g d (22) If the rise velocity is 1 fps and the drag coefficient is between 0.5 and 1.0, corresponding to a bubble shape somewhere between a sphere and a disk, the average size of broken up bubbles would be about 0.24 inches in diameter. These would be formed, according to Equation (21), a distance of less l than one inch after the wake forms and the initial bubble 15DA-41 49-B4
GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 15DA.3.2 Bubble Breakup Distance (Continued) reaches a corresponding terminal velocity. If the initial bubble rises between one or two radii before a wake forms, one expects sudden division into many small bubbles immediately after that. The hydrodynamic tests show a large bubble breaks away from its charging source, after which the lower surface appears to snap through to the top, shattering the entire bubble. This model simplifies the actual process by neglecting bubble interaction and incorporating the idealization of spherical bubbles with constant velocity and drag coefficients. However, it shows that even with energy dissipated by drag forces, and kinetic energy increase of the surrounding liquid, there is sufficient excess energy transfer to shatter the bubbles quickly. 15DA.3.3 Summary and Conclusions This analysis examined breakup mechanisms of gas bubbles rising through liquid. Free bubbles, which are at the average hydrostatic pressure, undergo breakup as they rise through liquid by buoyancy. Four bubble breakup mechanisms identified were: (1) inviscid flattening during which the lower surface overtakes the upper rising surface; (2) aerodynamic shredding, which pulls the bubble apart in a horizontal plane by higher liquid velocity past the bubble equator, correspondingly lower pressure, and a resulting outward force which overcomes surface tension; O 49-C56}}