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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20198K6611998-12-24024 December 1998 Safety Evaluation Supporting Amend 120 to License DPR-6 ML20154E0371998-09-30030 September 1998 Safety Evaluation Accepting Request for Exemption from Certain Portions of 10CFR50.47(b) & App E to 10CFR50 to Allow Brpnp to Discontinue Offsite EP Activities & Reduce Scope of Onsite EP as Result of Permanently Shutdown ML20154E0581998-09-30030 September 1998 Safety Evaluation Accepting Licensee Request from Exemption from Certain Portions of 10CFR50.47(b) ML20198K0091998-09-18018 September 1998 SER Accepting Licensee Request for Exemption from Certain 10CFR50 Requirements for Emergency Planning for Big Rock Nuclear Plant ML20216K0011998-04-16016 April 1998 Safety Evaluation Approving Licensee Request Re Plant Training Program for Certified Fuel Handlers ML20141J8731997-08-14014 August 1997 Safety Evaluation Supporting Amend 119 to License DPR-6 ML20137X0161997-04-18018 April 1997 Safety Evaluation Accepting Changes to Rev 17 of CPC Quality Program Description for Operational NPPs (CPC-2A) ML20137J9381997-04-0202 April 1997 Safety Evaluation Supporting Amend 118 to License DPR-6 ML20058F3441993-11-22022 November 1993 Safety Evaluation Concurring W/Contractor Findings Presented in Technical Evaluation Rept EGG-RTAP-10816, Evaluation of Utility Responses to Suppl 1 to NRC Bulletin 90-01;Big Rock Point ML20058A1601993-11-15015 November 1993 Safety Evaluation Supporting Amend 112 to License DPR-6 ML20057E1981993-10-0505 October 1993 Safety Evaluation Supporting Amend 111 to License DPR-6 ML20056E1661993-08-16016 August 1993 Safety Evaluation Supporting Amend 110 to License DPR-6 ML20128C9621992-11-27027 November 1992 Safety Evaluation Accepting Response to Suppl 1 to GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-46 ML20059H6051990-09-11011 September 1990 Safety Evaluation Approving Util 891229 Application for Disposal of Discharge Canal Dredging Spoils at Site ML20059F2581990-08-31031 August 1990 Safety Evaluation Approving Licensee Proposal to Dispose of Discharge Canal Dredgings Onsite in Manner Described in Util ML20246D2391989-08-16016 August 1989 Safety Evaluation Supporting Amend 100 to License DPR-6 ML20245G5211989-08-10010 August 1989 SER Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors NUREG-0123, Safety Evaluation Supporting Amend 99 to License DPR-61989-07-31031 July 1989 Safety Evaluation Supporting Amend 99 to License DPR-6 ML20245H8421989-07-28028 July 1989 Safety Evaluation Supporting Amend 98 to License DPR-06 ML20248C0621989-05-31031 May 1989 Safety Evaluation Supporting Amend 97 to License DPR-6 ML20246L8251989-05-0202 May 1989 Safety Evaluation Supporting Amend 96 to License DPR-6 ML20245F8391989-04-14014 April 1989 Safety Evaluation Supporting Amend 95 to License DPR-6 ML20235J0251989-02-15015 February 1989 Safety Evaluation Supporting Amend 94 to License DPR-6 ML20205T5911988-11-0404 November 1988 Safety Evaluation Supporting Requested Relief from Inservice Testing Requirements ML20205S1271988-10-14014 October 1988 Safety Evaluation Supporting Amend 93 to License DPR-6 ML20154G1131988-09-14014 September 1988 Safety Evaluation Supporting Amend 92 to License DPR-6 ML20154C1381988-09-0707 September 1988 Revised Safety Evaluation Accepting Continued Use of Hafnium Hybrid Control Blade & Proposed Surveillance Program ML20155F3511988-06-0606 June 1988 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification ML20154H4051988-05-17017 May 1988 Safety Evaluation Supporting Continued Use of Present Six Nucom Rods,Insertion of Two Similar Rods for Cycle 23 & Use of Surveillance Program ML20154H5341988-05-17017 May 1988 Safety Evaluation Supporting Amend 90 to License DPR-6 ML20154J1981988-05-17017 May 1988 Safety Evaluation Supporting Amend 91 to License DPR-6 ML20211P1411987-02-19019 February 1987 Safety Evaluation Supporting Issuance of Amend 89 to License DPR-6 ML20211N5401987-02-17017 February 1987 Safety Evaluation Supporting Issuance of Amend 88 to License DPR-6 ML20207S1681987-02-12012 February 1987 Safety Evaluation Concluding That Portions of Util 861205 Application to Amend License DPR-6,revising Tech Spec Section 5.2.1,Tables 1 & 2 Re Defining Operating Limits for New Reload I-2 Fuel Unacceptable ML20209H0651987-01-28028 January 1987 Safety Evaluation Supporting Amend 87 to License DPR-6 ML20212L9441987-01-16016 January 1987 Safety Evaluation Supporting Original Exemption from 10CFR50,App R Requirements Re Oil Collection Sys to Be Installed on Recirculation Pumps ML20198A3911986-05-12012 May 1986 Safety Evaluation Supporting Amend 85 to License DPR-6 ML20210P1761986-05-0606 May 1986 Safety Evaluation Supporting Amend 84 to License DPR-6 ML20155D7161986-04-11011 April 1986 Safety Evaluation Supporting Util 840730 Proposed Amend to License DPR-6,changing Tech Specs to Add Definition for Reportable Event & to Delete Specific Reporting Requirements Included in 10CFR50.72 & 50.73 ML20141N6571986-03-10010 March 1986 Safety Evaluation Supporting Amend 83 to License DPR-6 ML20154A1011986-02-12012 February 1986 Safety Evaluation Supporting Amend 82 to License DPR-6 ML20138K8001985-12-12012 December 1985 Safety Evaluation Supporting Util 850410 Request for Relief from Inservice Testing Requirements for Valves in Feedwater & Reactor Depressurization Nitrogen Backup Sys ML20136D1451985-11-19019 November 1985 Safety Evaluation Re Response to Generic Ltr 83-28,Items 3.1.1-3,3.2.1-3 & 4.5.1 Concerning post-maint & Reactor Trip Sys Functional Testing.Response Acceptable ML20138R2071985-11-15015 November 1985 Safety Evaluation Re Environ Qualification of Electric Equipment Important to Safety.Util Program Complies w/10CFR50.49 & Resolution of 830426 SER & Technical Evaluation Rept Acceptable ML20209J2401985-11-0505 November 1985 Safety Evaluation Supporting Util 831107 & 850816 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review Program & Description ML20198A9621985-11-0101 November 1985 Safety Evaluation Supporting Request for Relief from Inservice Insp Requirements ML20205F6051985-11-0101 November 1985 Safety Evaluation Supporting Amend 81 to License DPR-6 ML20205E9721985-10-29029 October 1985 Safety Evaluation Supporting Amend 80 to License DPR-6 ML20133N3931985-10-22022 October 1985 Safety Evaluation Supporting Amend 79 to License DPR-6 ML20137W3231985-10-0202 October 1985 Safety Evaluation Supporting Amend 78 to License DPR-6 1998-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217C3031999-09-28028 September 1999 Annual Rept of Facility Changes,Tests & Experiments ML20199A6621999-01-0505 January 1999 Special Rept:On 981230,hi Range Noble Gas Monitor Was Inoperable for Greater than Seven Days.Cause Unknown. Preplanned Alternate Method of Monitoring Appropriate Parameters within 72 H Was Established ML20206F6131998-12-31031 December 1998 1998 Consumers Energy Co Annual Rept. with ML20198K6611998-12-24024 December 1998 Safety Evaluation Supporting Amend 120 to License DPR-6 ML20154E0371998-09-30030 September 1998 Safety Evaluation Accepting Request for Exemption from Certain Portions of 10CFR50.47(b) & App E to 10CFR50 to Allow Brpnp to Discontinue Offsite EP Activities & Reduce Scope of Onsite EP as Result of Permanently Shutdown ML20154E0581998-09-30030 September 1998 Safety Evaluation Accepting Licensee Request from Exemption from Certain Portions of 10CFR50.47(b) ML20198K0091998-09-18018 September 1998 SER Accepting Licensee Request for Exemption from Certain 10CFR50 Requirements for Emergency Planning for Big Rock Nuclear Plant ML20217N2131998-04-24024 April 1998 Brpnp Zircaloy Oxidation Analysis ML20216K0011998-04-16016 April 1998 Safety Evaluation Approving Licensee Request Re Plant Training Program for Certified Fuel Handlers ML20217H4641998-03-26026 March 1998 Rev 2 to Post Shutdown Decommissioning Activities Rept (Psdar) ML20202G1941998-02-12012 February 1998 Rev 7 to Updated Final Hazards Summary Rept for Big Rock Point Plant ML20154A7591997-10-0808 October 1997 10CFR50.59 Annual Rept of Facility Changes,Tests & Experiments, Since 971008 ML20216E4731997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Big Rock Point Plant ML20141J8731997-08-14014 August 1997 Safety Evaluation Supporting Amend 119 to License DPR-6 ML20210H5601997-07-31031 July 1997 Monthly Operating Rept for July 1997 for Brpnp ML20148T4901997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Big Rock Point Nuclear Plant ML20148N9251997-06-0606 June 1997 Rev 18 to CPC-2A, Quality Program Description for Operational Nuclear Plants ML20140C8981997-05-31031 May 1997 Monthly Operating Rept for May 1997 for Big Rock Nuclear Power Plant ML20138J0121997-04-30030 April 1997 Monthly Operating Rept for Apr 1997 for Big Rock Point ML20137X0161997-04-18018 April 1997 Safety Evaluation Accepting Changes to Rev 17 of CPC Quality Program Description for Operational NPPs (CPC-2A) ML20137J9381997-04-0202 April 1997 Safety Evaluation Supporting Amend 118 to License DPR-6 ML20137P0391997-03-31031 March 1997 Monthly Operating Rept for Mar 1997 for Big Rock Point Nuclear Plant ML20135F2361997-02-28028 February 1997 Monthly Operating Rept for Feb 1997 for Big Rock Nuclear Plant ML20148N9181997-02-0101 February 1997 Rev 17 to CPC-2A, Quality Program Description for Operational Nuclear Plants ML20134H3691997-01-31031 January 1997 Monthly Operating Rept for Jan 1997 for Big Rock Point Nuclear Plant ML20137F2101996-12-31031 December 1996 1996 Annual Financial Rept CMS Energy ML20133C5421996-12-31031 December 1996 Monthly Operating Rept for Dec 1996 for Big Rock Point Nuclear Plant ML20135E5101996-11-30030 November 1996 Monthly Operating Rept for Nov 1996 for Big Rock Point Nuclear Plant ML20134H3381996-10-31031 October 1996 Monthly Operating Rept for Oct 1996 for Big Rock Point Nuclear Plant ML20211N1561996-10-0808 October 1996 Annual Rept of Facility Changes,Tests & Experiments, Consisting of Mods & Miscellaneous Changes Performed Since 961008 ML20128F9821996-09-30030 September 1996 Monthly Operating Rept for Sept 1996 for Big Rock Point Nuclear Plant ML20059E8321993-12-31031 December 1993 Monthly Operating Rept for Dec 1993 for Big Rock Point Nuclear Plant ML20058K0931993-11-30030 November 1993 Monthly Operating Rept for Nov 1993 for Big Rock Point Nuclear Plant ML20058E8961993-11-29029 November 1993 1993 ISI Rept 3-1 Big Rock Point Plant, for 930626-0905 ML20058F3441993-11-22022 November 1993 Safety Evaluation Concurring W/Contractor Findings Presented in Technical Evaluation Rept EGG-RTAP-10816, Evaluation of Utility Responses to Suppl 1 to NRC Bulletin 90-01;Big Rock Point ML20058G5981993-11-17017 November 1993 Part 21 Rept Re Westronics Recorders,Model 2100C.Signal Input Transition Printed Circuit Board Assembly Redesigned to Improve Recorder Immunity to Electromagnetic Interference.List of Affected Recorders & Locations Encl ML20058A1601993-11-15015 November 1993 Safety Evaluation Supporting Amend 112 to License DPR-6 ML20059J4531993-10-31031 October 1993 Monthly Operating Rept for Oct 1993 for Big Rock Point Nuclear Plant ML20057G1511993-10-0707 October 1993 Part 21 Rept Re Westronics Model 2100C Series Recorders. Informs That Over Several Tests,Observed That Recorder Would Reset During Peak Acceleration & Door Being Forced Off Recorder.Small Retaining Clips Added to Bottom of Door ML20057E1981993-10-0505 October 1993 Safety Evaluation Supporting Amend 111 to License DPR-6 ML20057E8341993-09-30030 September 1993 Monthly Operating Rept for Sept 1993 for Big Rock Point Nuclear Plant ML20056G9171993-08-31031 August 1993 Monthly Operating Rept for Aug 1993 for Big Rock Point Nuclear Plant ML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20056E1661993-08-16016 August 1993 Safety Evaluation Supporting Amend 110 to License DPR-6 ML18058B8821993-06-15015 June 1993 Rev 13 to Quality Program Description for Operational Nuclear Power Plants. ML20128P5501993-02-18018 February 1993 Section 2.5 of Big Rock Point Updated Final Hazards Summary Rept ML20128F3511993-01-31031 January 1993 Monthly Operating Rept for Jan 1993 for Big Rock Point Nuclear Plant ML20128C4341993-01-29029 January 1993 Forwards Rev 3 to Updated Final Hazards Summary Rept ML20058L8721992-12-31031 December 1992 1992 Annual Rept,Cpc ML20127K8941992-12-31031 December 1992 Revised Pages to Graybook Rept for Dec 1992 for Big Rock Point Nuclear Plant 1999-09-28
[Table view] |
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4 UNITED STATES ATOMIC ENERGY COMMISSION SAFETY EVALUATION BY THE DIRECTORATE OF LICENSING DOCKET No. 50-155 BIG ROCK POIhT PLANT ANALYSIS OF THE CONSEOUENCES OF HIGH ENERGY PIPING FAILURES OUTSIDE CONTAINMENT INTRODUCTION On December 18, 1972, and January 16, 1973, the Atomic Energy Commission's Regulatory staff sent letters to Consumers Power Com'pany requesting a detailed design evaluation to substantiate that the design of the Big Rock Point Plant is adequate to withstand the effects of a postulated rupture in any high energy fluid piping system outside the primary containment, including the double-ended rupture of the largest line in the main steam and feedwater system. It was further requested that if the results of the evaluation indicated that changes in the design were necessary to assure safe plant shutdown, information on these design changes and plant modifications would be required.
Criteria for conducting this evaluation were included in the letters. A meeting was held on January 24, 1973, to discuss the information already available on the Big Rock Point Plant design concerning postulated pipe ruptures, to discuss the criteria, and to assess those areas where additional information was required. In response to our letters, a report c o n c e r n ir.g postulated high energy pipe ruptures outside containment was filed by Consumers Power on June 29,'1973. A subsequent letter from Consumers Power dated February 7, 1974, answered additional questions in a letter from the staff dated January 23, 1974.
EVALUATION .
Criteria A summary of the criteria and requirements in.cluded in our letter of December 18, 1972, is set forth below:
- a. Protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown c.o n d i t i o n ,
assuming a concurrent and unrelated single active failure of protected equipment, should be provided from all effects I
f3L_~fi b9!
.i o
' ~
2-resulting from ruptures in pipes carrying high energy.
fluid, where the temperature and pressure conditions of the.
fluid exceed 200*F and 275 psig, respectively, up to and including a double-ended rupture of such pipes. Breaks should be' assumed to occur in those locations specified in the " pipe whip criteria". The rupture effects to be considered include. pipe whip, structural (including the effects of jet' impingement), and environmental.
- b. In addition, protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protected equipment, should be provided from.the environmental and structural effects (including the - effects of jet impingement) resulting from a single open crack at.
the most ~dverse a location in pipes carrying fluid routed
'i in the vicinity of this e,quipment. 'The size of the cracks should be assumed to be 1/2 the pipe diameter-in length and 1/2 the wall thickness in width.
High Energy Systems Our evaluation included the following piping systems contair,ing high energy fluids: ,
Main, Extraction, and Auriliary Steam Systems Feedwater System Condensate System Sample Lines (Environmental Effects Only)
Areas or Systems Affected by High Energy Pipe Breaks-An evaluation was conducted of the effects of high energy pipe breaks on the following systems, components, and structures which ,
would be necessary (in various combinations, depending.on the l effects of the break) to safely shutdown, cooldown, and maintain ;
cold shutdown conditions: i
- a. General
.1. Control Room ,
- 2. Control and Instrument' Cables and' Tunnels
, . . _ , . , . , . , , , . - - . _ . . - , . , , , ~ . - . . _ , , - - - - _ . . - -,,,..- . - . - . _ ~ . . . . _ _ . _ _ . . _ . . _ . .-
i
- 3. Electrical Distribution System
- 4. Fmergency de Power Supply (batteries) .
Emergency ac Power Supply (diesels) j 5.
I
- 6. Heating and Ventilation Systems (needed for long-term occupancy *o maintain the reactor in safe shutdown condition)
- b. Reactor Control Systems and associated instrumentation
- c. Cooling and Service Water Systems
- d. ECCS components Specific Areas of Concern The applicant has provided the results of his examination of all postulated safety related high energy line break locations and evaluated the break consequences. We have reviewed all of this information, including the following specific areas of concern where the potential consequences might be severe or where specific corrective action, would further assure safe cold shutdown of the plant.
- a. Compartment Pressurization Large pipe breaks, including the double-ended rupture of the largest pipes in a system, and small leakage cracks up to the design basis size have been considered for the main steam tunnel, the turbine building, and the electrical penetration room.
in the main steam tunnel, the effects of a combined main i steam line break and a feedwater line rupture were considered as the worst case. The existing pipe tunnel and attached condenser area structure are abic to resist loads due to pressurization of 2.4 psi with modifications. The resultant l
pressure from a piping failure inside the tunnel was calculated I to increcse to 20 psi. Modifications of the existing structures wcTe considered to assure the integrity of the tunnel in the
. event of a postulated failure. To reduce the pressure below the 2.4 psi, approximately 140 square feet of additional vent area would be required. Since there is a ventilation connection
.\
between the tunnel and' adjacent electrical penetration room, a fast closure damper will be installed to mitigate environ-mental effects in the electrical penetration room.
- b. Pipe Whip 1
The steam tunnel has been designed with thick reinforced
. concrete capable of withstanding large static and dynamic loads. The reinforced concrete steam tunnel in which the main steam and feedwater lines are routed from the prima'ry containment to the turbine room is subjected to the loads of the piping and a live load from the floor on top of the tunnel roof. The possibility of a combined whip, jet thrust and pressure pulse acting upon the tunnel wall and on the containment was evaluated by the applicant. Break points of the lines were selected using the criteria in our
- December 1972 and January 1973 letters. A whipping line could crack the tunnel walls or overhead, but would only cause spalling and therefore would not generate any signi-ficant missiles which would damage safety related equipment.
A whipping steam or feedwater line could cause the 4 and 6 inch core spray headers in the pipe tunnel to be severed.
- However, a redundant supply is available through a separate penetration approximately 180' around the containment from >
the pipe tunnel.
Other high energy lines such as the sample lines and reactor water cicanup line are located such that their rupture would not cause damage to -the safety related equipment.
- c. Control Room Habitability .
l ,
The main control room is physically located away from and isolated from all high energy lines. Neither the control room equipment nor its ventilation system will be affected by environmental effects caused by a rupture of a high energy line.
- d. Environmental Effects Components and equipment were checked for possible adverse environmental effects which could be caused by the rupture of a high energy line. Adverse temperature, pressure, and humidity were the parameters which were used in the evaluation of safety related equipment.
- i. n.
We have reviewed th'e licensee assessment of the consequences of environmental ef fects on safety related equipment. We find that safety related equipment has been designed to limits in excess of postulated conditions which could arise from the rupture of a high energy line.
Modifications Modifications to the existing facility will be required to assure that the design will have adequate safety margins in the event of a high energy line rupture outside the containment.
Modifications to existing structures are necessary to assure the integrity of the main steam tunnel and surrounding areas.
To reduce the pressure below 2.4 psi in the main steam tunnel, 130 to 140 square feet of additional vent area is required. Two acceptable alternatives were coesidered:
- a. A vent hole in the west side of the pipe tunnel between the reactor and turbine buildings,
- b. A vent hole in the concrete block shield wall on the west side of the condenser.
In addition to the vent opening, a ventilation duct between the pipe tunnel and the electrical penetration room will be modified to provide an automatic isolation in the event of tunnel pressurization.
CONCLUSIONS On the basis of this review of ti e information submitted and our discussions with Consumers Power, we find that their assessment of the consequences of high energy line failures outside containment is acceptabic. Some modifications are necessary. We have concluded that the potential consequences of the postulated high energy pipe failures, following the modifications, will not prevent the capability to achieve safe cold shutdown conditions consistent with the single failure and redundancy requirements as described in our letter of December 18, 1972.
i 4
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The licensee has stated by telecon that modifications will not ,
be complete until July 1975. In the interim we will require the licensee to undertake an increased surveillance program in'the '
areas of concern.
Based on the decreased likelihood of a catastrophi'c high energy line break, owing to the increased surveillance and the limited
' time until the final modifications will be completed, we have i concluded that there is reasonable assurance that the healtl) and safety of the public will not be endangered by continued I operation in the manner proposed. t i
b ' g7%b od ames C. Snell perating Reactors Branch #2 '
Directorate of Licensing iO ppiu. .
'I
% wrd Dennis L. Ziemann, Chief Operating Reactors Branch #2 Directorate of Licensing :
Date: 3 gl, f
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