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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20216F0221998-03-0606 March 1998 Safety Evaluation Accepting Request Re Temporary Use of Current Procedure for Containment Repair & Replacement Activities at Plant ML20197B9171997-07-23023 July 1997 Safety Evaluation Re Concrete Expansion Anchor Safety Factors for High Energy Line Break Restraints ML20141E5091997-05-16016 May 1997 Safety Evaluation Supporting TR EMF-96-051(P), Application of Anfb Critical Power Correlation to Coresident GE Fuel for Plant,Unit 2 Cycle 15 ML20137G6071997-03-13013 March 1997 Safety Evaluation Supporting Proposed Changes to TS & Bases Ceco ML20134H7601997-02-0707 February 1997 Safety Evaluation Approving Rev 65c of Ceco QA TR CE-1-A ML20149F4151994-08-0404 August 1994 Safety Evaluation Concluding That Unit 1 Can Be Safely Operated During Next Operating Cycle (Cycle 14) ML20058L2711993-12-0808 December 1993 Safety Evaluation Finding Overlay Repair of Weld 02C-F7 Acceptable & in Conformance W/Gl 88-01.Plant May Be Returned to Safe Operation ML20056C4601993-06-17017 June 1993 Safety Evaluation Accepting Proposed Repair of Weld in Recirculation Piping Sys for One Cycle of Operation ML20128F9731993-02-10010 February 1993 Safety Evaluation Granting Licensee 910930 Request Not to Perform Code Exam on 100% of Attachment Welds on Stabilizer Brackets to Reactor Vessel Under 10CFR50.55(a)(3)(ii) ML20055F9221990-07-17017 July 1990 Safety Evaluation Supporting Util Responses to NRC Bulletin 88-010 Re Molded Case Circuit Breaker Replacement ML20248J2431989-10-0303 October 1989 Safety Evaluation Accepting Util 880122,0601,0714 & 0816 Submittals Re Insp Results,Mitigation,Flaw Evaluations & Overlay Repairs of Welds Susceptible to IGSCC to Support Operation of Unit 2,for Another 18-month Fuel Cycle ML20246K1611989-08-24024 August 1989 Revised SER Supporting Amends 112 & 108 to Licenses DPR-29 & DPR-30,respectively,changing Setpoints of Main Steam Line Radiation Monitors & Correcting Typos in Tech Specs ML20248B8911989-06-0606 June 1989 Safety Evaluation Concluding That IGSCC Insp Scope for Class 1 Piping Meets NRC Requirements & Guidelines of Generic Ltr 84-11 ML20151X3431988-08-16016 August 1988 SER Accepting Basis & Findings That Util post-accident Monitoring Instrumentation Meets Guidelines of Reg Guide 1.97 Except for Variable Neutron Flux Instrumentation ML20151M6901988-07-21021 July 1988 Revised Safety Evaluation Supporting Exemption Requests from Regulatory Requirements of 10CFR50,App R,Section Iii.G ML20195E2091988-06-0909 June 1988 Safeguards Evaluation Rept Supporting Amends 108 & 103 to Licenses DPR-29 & DPR-30,respectively ML20151U1201988-04-20020 April 1988 Revised Safety Evaluation Accepting Util Interim Compensatory Measures & Request for Exemption from 10CFR50, App R,Section Iii.G Requirement Re Hot Shutdown Repair for Fire Event in Plant ML20149M5301987-12-11011 December 1987 Marked-up Safety Evaluation Supporting Request for Exemptions from App R ML20236W4851987-12-0101 December 1987 Safety Evaluation Accepting Proposed Approaches for Resolving fire-related Concerns,Including Spurious Operations,High Impedance Faults & Electrical Isolation Deficiency.Granting of Exemption Requests Recommended ML20235S8541987-10-0202 October 1987 Safety Evaluation Supporting Interim Approval of Rev 3 to Process Control Program for Plant ML20237H7061987-08-19019 August 1987 SER Supporting Util Response to Item 2.1 (Part 1) of Generic Ltr 83-28 Re Equipment Classification.Licensee Statements Confirm Program Exists for Identifying,Classifying & Treating Components as safety-related.Program Acceptable ML20236H1341987-07-27027 July 1987 Safety Evaluation Re Acceptance of Updated Rev 11 to Offsite Dose Calculation Manual ML20205H1351987-03-23023 March 1987 Safety Evaluation Re Insps for & Repairs of Igscc.Facility Can Be Safely Operated for One 18-month Fuel Cycle in Present Configuration ML20214X1111986-11-26026 November 1986 Safety Evaluation Supporting Util Analytical Methods Used to Evaluate Stresses of Critical Components for Vacuum Breaker Integrity Re Mark I Containment Program ML20214Q3851986-11-17017 November 1986 Safety Evaluation Re Insp & Repair of Reactor Coolant Piping Sys ML20141D2291986-03-31031 March 1986 Safety Evaluation Granting Util Request for Relief from Certain Requirements of Section XI of ASME Code Re Inservice Insp for Second 10-yr Interval ML20141P0491986-03-13013 March 1986 Safety Evaluation Supporting Licensee 831105 & 851219 Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review (Data & Info Capability) ML20137A3931986-01-0707 January 1986 Safety Evaluation Supporting Reactor Coolant Piping Sys IGSCC Insp & Repair Per Generic Ltr 84-11 & Return to Operation for 18-month Cycle ML20133F0291985-07-30030 July 1985 Safety Evaluation Accepting Util 831105 & 850605 Responses to Generic Ltr 83-28,Item 1.1 Re post-trip Review (Program Description & Procedure) ML20126F4561985-05-31031 May 1985 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1 Re post-maint Testing Verification ML20062B8351982-07-28028 July 1982 Safety Evaluation Supporting Plant Compliance W/Esf Reset Controls Per NRC Criteria ML20126C3461980-03-20020 March 1980 Safety Evaluation Supporting Amend 51 to License DPR-30 ML20235D0971966-12-30030 December 1966 Safety Evaluation Supporting Util 660531 Proposal to Const & Operate Single Cycle BWR of 2,255 Mwt 1999-09-21
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data SVP-99-204, Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 SVP-99-179, Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs SVP-99-155, Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-148, Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety SVP-99-123, Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations SVP-99-104, Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-102, Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with1999-04-30030 April 1999 Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-071, Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20205C5671999-03-19019 March 1999 Simulator Four-Yr Certification Rept ML20207D2341999-03-0101 March 1999 Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205 ML20204B1571999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.With SVP-99-021, Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With1999-01-31031 January 1999 Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With ML20205D1311998-12-31031 December 1998 1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with SVP-99-007, Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With ML20196C8391998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-030-2, Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1 SVP-98-364, Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196G1241998-11-30030 November 1998 COLR for Quad Cities Unit 1 Cycle 16 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196C8731998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-30-1, Fracture Mechanics Evaluation on Observed Indications at Two Welds in Recirculation Piping of Quad Cities,Unit 1 Station ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB SVP-98-346, Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-98-358, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With1998-10-31031 October 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With SVP-98-326, Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20153D0191998-09-18018 September 1998 Part 21 Rept Re Defect in Gap Conductance Analyses for co- Resident BWR Fuel.Initially Reported on 980917.Corrective Analyses Performed Demonstrating That Current Operating Limits Bounding from BOC to Cycle Exposure of 8 Gwd/Mtu ML20153C6771998-09-17017 September 1998 Part 21 Rept Re Defect Relative to MCPR Operating Limits as Impacted by Gap Conductance of co-resident BWR Fuel at Facilities.Operating Limit for LaSalle Unit 2 & Quad Cities Unit 2 Will Be Revised as Listed ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20151Y7261998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Quad Cities Nuclear Power Station ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20151Y7301998-07-31031 July 1998 Revised MOR for Jul 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20237A6251998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Quad Cities Nuclear Power Station,Unit 1 & 2 SVP-98-328, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With1998-07-15015 July 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With SVP-98-249, Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-98-215, Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 2 ML20247N6281998-05-19019 May 1998 Rev 2 to COLR for Quad Cities Unit 2 Cycle 15 ML20216C0561998-04-30030 April 1998 Safe Shutdown Rept for Quad Cities Station,Units 1 & 2, Vols 1 & 2.W/22 Oversize Figures SVP-98-176, Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20217D0281998-04-22022 April 1998 Part 21 Rept Re Additive Constants Used in MCPR Determination for Siemens ATRIUM-9B Fuel by Core Monitoring Sys Were Found to Be non-conservative.SPC Personnel Notified All Customers w/ATRIUM-9B Lead Test Assemblies ML20217G3951998-04-0808 April 1998 TS 3/4.8.F Snubber Functional Testing Scope Quad Cities Unit 2 TS (Safety-Related) Snubber Population 129 Snubbers SVP-98-128, Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 2 1999-09-30
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'p <[pcsc ^ UNITED STATES 18 w NUCLEAR REGULATORY COMMISSION 7, WASHINGTON, D. C. 20555 T, ..... f;E I
SAFETY EVALUATION BY THE'0FFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 112 TO FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 108 TO FACILITY OPERATING LICENSE NO. DPR-30 l
COMMONWEALTH EDISON' COMPANY AND l
IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letters dated September 16 and November the licensee requested an amendment to Quad Cities Station 18,(1988, Units 1 and 2) Operating
' Licenses DPR-29 and DPR-30 to change the setpoint of main steam line radiationmonitors(MSLRMS)andcorrecttypographicalerrorsinthe
-Technical Specifications. The requested change involves increasing the setpoint of MSLRMs from seven times Normal Full Power Background (NFPB) to 15 times NFPB (without hydrogen addition) to allow for implementation of Hydrogen Water. Chemistry (HWC) which is expected to mitigate the effectsofIntergranularStressCorrosionCracking(IGSCC). The MSLRM setpoint change is necessary since the injection of hydrogen into the feedwater lowers the oxidizing potential in the reactor coolant which in turn converts more N-16 to a volatile species and results in an
. increase in steam line radiation level. As a consequence, the NFPB sta:am activity during hydrogen addition can increase up to approximately a factor of five times greater than NFPB steam activity without hydrogen addition.
By letters dated September 28, 1988 and May 1, 1989, the licensen provided additional information to support the implementation of HWC. The additional information included:
(1) Report titled, "HWC Installation Report for Amendment to the facility Operating License," dated May 16, 1988.
(2) Report titled, "HWC Installation Compliance with Electric Power Research Institute (EPRI) Guidelines for Permanent BWR Hydrogen Water Chemistry Installations - 1987 Revision."
(3) Draft copy of Proposed Changes to Updated FSAR as a Result of HWC Addition at Quad Cities Station.
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The changes will be included in the June 30, 1989 update to the FSAR.
l 2.0 EVALUATION 2.1 MSLRM Setpoint The MSLRMs provide reactor stram and main steam line isolation signals when high-activity levels are detected in the main steam lines. Addi-tionally, these monitors serve to limit radioactivity releases in the event of fuel failures. Technical Specification (TS) changes are needed to accommodate the expected main steam line radiation levels (for increased N-16 activity levels in the steam phase) as a result of hydrogen injection into the reactor coolant system.
The licensee has requested TS changes to raise MSLRM set points from the current seven times NFPB to fif teen times NFPB (without hydrogen addition). The licensee proposes a single set point for the MSLRMs which is an exception to the EPRI " Guidelines for Permanent BWR Hydrogen Water Chemistry Installations - 1987 Revision" (hereafter referred to as the Guidelines). The Guidelines recommended a dual MSLRM set point: (1) For reactor power less than 20% of rated, when hydrogen should not be injected, the setpoint is maintained at the current TS factor above NFPB, and (2) For reactor power greater than 20% of rated, the set point is readjusted to the same TS factor above NFPB with hydrogen addition.
The only design basis event in which the Quad Cities Station takes credit for th? MSLRM is the Control Rod Drop Accident (CRDA). In the event of a CRDA, the MSLRMs detect high radiation levels in the main steam lines and provide signals for reactor scram and Main Steam Line Isolation Valve (MSIV) closure to reduce the release of fission products to the environment. For the proposed MSLRM set point of fifteen times NFPB (without hydrogen addition), the calculated dose rate at the MSLRM is 1.5 R/hr. For a CRDA, the dose rate at the MSLRM is 8R/hr. Since the MSLRM dose rate from the CRDA is over five times the proposed increased MSLRM set point, the high radiation signal caused by the CRDA will still scram the reactor and isolate the MSIVs.
Raising the MSLRM trip set point from the current 0.7 R/hr to 1.5 R/hr will not result in a significant increase in the radiological consequences of a CRDA. The time to reach the proposed MSLRM trip set point following a CRDA will be increased by less than 1/4 second. The Quad Cities TS permits five seconds for MSIV closure. The incre se in time-to-closure due to the proposed MSLRM set point is only 5% of the current time-to-closure. Since the calculated dose from the CRDA is only 12 mrem, the minor increase in MSIV isolation will have an insignificant effect on the total activity release and resulting dose to the general public.
In the event of an incident causing minor fuel damage such that radiation levels will not exceed the proposed MSLRM set point of fifteen times NFPB (without hydrogen addition), the downstream steam jet air ejectors radiation detectors would be alarmed. These detectors have a greater sensitivity thantheMSLRMsfornoblegasesbecauseoftheholdupperiod(delay
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betweenMSLRMandsteamjetairejectorradiationdetectors)whichallows for significant decay of H-16 (7.1 second half-life). Since steam jet air ejector radiation detectors are in the Quad Cities (Units 1 and 2) TS, the proposed MSLRM set point change will not result in offsite doses in excess of established release limits.
On the basis of the above evaluation, we find that the proposed Technical Specification changes required for implementation of HWC at Quad Cities Station (Units 1 and 2) are acceptable. The proposed increased single set point, versus a dual power dependent set point, for the MSLRMs is an exception to the BWR Owners Group, " Guidelines for Permanent Hydrogen Water Chemistry Installations - 1987 Revision." This exception is justified on the basis that the CRDA dose rate is already limiting at five times the new set point. Thus, it will not affect the safety of the plant or the general'public.
2.2 Radiation Protection The staff has reviewed the licensee's submittal regarding the radiological implications due to the increased dose rate associated with increased N-16 activity levels during hydrogen injections into the reactor system. The licensee is committed to designing, installing, operating, and maintaining the HWC System in accordance with Regulatory Guides 8.8 and 8.10 to assure that occupational radiation exposures and doses to the general public will be As Low As Reasonably Achievable (ALARA). A preliminary radiological study has been completed at the Quad Cities Station to identify areas of the station which may experience increased dose rates due to HWC. When HWC is implemented, the results of the preliminary study will be confirmed and additional measurements will be made, if required. Based on the preliminary HWC study)and in March experience 1983 , additional from appears shielding the Dresden to be Unit 2 (implemented unnecessary. Again, when HWC is implemented, these results will be confirmed and additional shielding will be provided, if required.
Plant procedures will aadress access control of radiation arens that are affected by HWC. Guidelines will be established for any additional controls needed for area posting and monitoring due to HWC. The existing radio-logical surveillance program (Section 8.4 of Offsite Dose Calculational Manual) assures compliance with regulatory requirements for offsite doses to the public. The licensee developed a temporary procedure for an extensive environmental monitoring program with an additional 70 TLDs. The program appears acceptable in determining radiation level increases from N-16 due to HWC.
Electrochemical Potential (ECP) measurements will be made to assure IGSCC protection as well as optimizing hydrogen injection rates to minimize radiation levels. Radiation protection practices implemented for HWC will ensure ALARA in accordance with Regulatory Guide 8.8 and are, there-fore, acceptable.
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2.3 Hydrogen an.10xygen Storage Facilities The hydrogen storage facility contains a liquid hydrogen tank, gaseous storage tubes, and two gaseous tube trailer discharge stations. The licensee will utilize a 20,000 gallon liquid hydrogen tank as a long term hydrogen source. Gaseous hydrogen storage tubes (total capacity 50,000 - 75 000 scf, each tube capacity 8,300 scf, maximum pressure 2,400 psig),are provided to serve as a gaseous surge volume for the liquid hydrogen tank. If the liquid hydrogen system is not completed, the licensee will utilize two gaseous hydrogen tube trailers for initial startup and operation. Gaseous tube trailers will also be brought onsite to provide backup hydrogen supply when liquid hydrogen is not available.
The pressure control station has two parallel full flow pressure reducing regulators. An excess flow check valve is installed downstream of the interim tube trailer and long-term liquid hydrogen storage facility. An additional excess flow check valve is installed in the hydrogen gas supply ifne near the west wall of the Unit 1 turbine building. Each excess flow check valve has a stop-flow-setpoint of 200 scfm (plant's hydrog'en flow requirements are 140 scfm).
The liquid hydrogen tank is constructed in accordance with Section VIII, Division 1 of the ASME Code for Unfired Pressure Vessels. The hydrogen storage facility (compressed gas and liquid) is located 1500 feet from the nearest safety-related structure. This distance meets the Guidelines which require 140 and 962 feet separation distance in the event of an explosion of a gaseous hydrogen storage tube and liquid hydrogen tank, respectively.
The hydrogen supply facility provides the gaseous hydrogen requirements for turbine generator cooling / purging as well as HWC for Units 1 and 2.
The liquid oxygen storage tank, with a maximum capacity of 11,000 gallons, is located 1000 feet away from the nearest safety-related air intake.
The hydrogen and oxygen storage facilities meet the Guidelines.
2.4 Hydrogen and Oxygen Injection System Hydrogen )iping is run underground from the storage facility to the outer wall of t,e Unit 1 turbine building. This piping is covered with a protective coating to protect against corrosion and is electrically grounded. The hydrogen injection lines for each unit are equipped with check valves and solenoid isolation valves which are interlocked with the condensate pump. Individual solenoid isolation valves provide hydrogen flow isolation if the associated condensate pump is shut down and for all hydrogen injection trips. Hydrogen is injected into the condensate pump discharge (at a rate of 70 SCFM at full power) to provide adequate dissolving and mixing and to avoid gas pockets at high points. The hydro-gen injection rate is automatically programmed as a function of main steam flow. Hydrogen injection is automatically tripped on main steam flow of less than 20%.
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The HWC system is tripped by the following signals:
Reactor scram, Low residual off-gas oxygen concentration, High hydrogen flow, Low hydrogen flow, Area hydrogen concentrator. high, Operator manual, Hydrogen storage facility trouble, and Low reactor steam flow Each unit has eight hydrogen area monitors located in the vicinity of hydrogen injection system components that may leak. The sensors feed to a monitor panel which trips the respective unit's HWC system at 20% of the lower explosive limit.
Oxygen is injected into the off-gas system to insure that all excess hydrogen in the off-gas stream is recombined. At the present time, oxygen is not piping into the condensate system to maintain recomended dissolved oxygen concentrations of 20-60 ppb for feedwater pipe corrosion control.
Should oxygen concentrations decrease below 20 ppb when HWC is implemented, the licensee comitted to provide for oxygen addition into the condensate system.
The hydrogen and oxygen injection system meet the Guidelines.
3.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that these amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such findings. Accordingly, these amendments meet the eligibility criteria forcategoricalexclusionsetforthin10CFR51.22(c)(9)and10CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of these amendments.
4.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, endsuch (2) public activities will be conducted in compliance with tne Comission's regulations
, and the 4suance of these amendments will r.ot be inimical to the common defense and security nor to the health and safety of the public.
l i Principal Contributor: Frank Witt Dated: August 24, 1939 L_- _