ML20236G266
| ML20236G266 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 07/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1002, NUREG-1002-S04, NUREG-1002-S4, NUDOCS 8708040187 | |
| Download: ML20236G266 (22) | |
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Supplement:No.f4 Safety Evaluation Report relateditoltheLoperation of Braidwood Station, LUriits 1 and 2 Docket Nos. STN'50-456 and STN 50-457 q
Commonwealth Edison. Company' U.S. Nuclear Regulatory 1
Commission Office of Nuclear Reactor Regulation July 1987 1
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l NOTICE l
Availability of Reference Materials Cited in NRC Publications f
I Most documents cited in NRC publications will be available from one of the following sources:
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- 1. The NRC Public Document Room,1717 H Street, N.W.
j Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082,.
Washington, DC 20013-7082 i
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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-
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. ment Room include NRC correspondence and internal NRC memoranda; NRC Office of inspection
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NUREG-1002 Supplement No. 4 Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 Docket Nos. STN 50-456 and STN 50-457 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1987 pe=
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ABSTRACT In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457).
The first supplement to NUREG-1002 was issued in September 1986; the second supplement was issued in October 1986; and the third supplement was issued in May 1987.
This fourth supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 3 was published.
The facility is located in Reed Township, Will County, Illinois.
Braidwood SSER 4 iii
TABLE OF CONTENTS P_ag iii ABSTRACT...............................................................
1 INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY..................
1-1 1.1 Introduction.................................................
1-1 1-1 1.7 Summary of Outstanding Items.................................
1-3 1.8 Confirmatory Issues.........................................
- 1. 9 License Conditions...........................................
1-5 5
REACTOR C0OLANT SYSTEM............................................
5-1 5.2 Integrity of Reactor Coolant Pressure Boundary...............
5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing................................
5-1 5.2.4.6 Evaluation of Compliance With 10 CFR 50.55a(a)(3) for Braidwood Unit 2.....
5-1 6
ENGINEERED SAFETY FEATURES........................................
6-1 6.4 Control Room Habitability....................................
6-1 6.5 Fission Product Removal and Control System...................
6-2 6.5.1 Engineered Safety Feature Atmospheric Cleanup System..
6-2 9
AUXILIARY SYSTEMS.................................................
9-1 9.5 Other Auxiliary Systems......................................
9-1 9.5.1 Fire Protection Program...............................
9-1 9.5.1.5 Fire Protection for Specific Plant Areas.....
9-1 9.5.4 Emergency Diesel Engine Fuel Storage and 9-1 Transfer System 9.5.4.2 Emergency Diesel Engine Fuel Oil Storage and Transfer System Design.......................
9-1 18 HUMAN FACTORS ENGINEERING.........................................
18-1 18.2 Main Control Room and Remote Shutdown Panel..................
18-1 18.3 Safety Parameter Display System..............................
18-1 Braidwood SSER 4 v
TABLE OF CONTENTS (Continued)
APPENDICES APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 APPENDID B BIBLIOGRAPHY APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS I
Braidwood SSER 4 vi
r 1 IN1RODUCTION AND GENERAL DESCRIPTION OF FACILITY 1.1 Introduction In November 1983, the Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) (NbdEG-1002) on the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457).
At that time, the staff identified items that had not been resolved with the applicant.
The first supplement to NUREG-1002 was issued in Septem'oer 1986; the second supplement to NUREG-1002 was issued in October 1986; and the third supplement to NUREG-1002 was issued in May 1987.
The purpose of this fourth supplement to the SER is to provide the staff evaluation of the open items that have been resolved to date and to address changes to the SER that resulted from the re-ceipt of additional information from Commonwealth Edison Company (licensee).
Each of the following sections or appendices is numbered the same as the corres-ponding SER section or appendix that is being updated.
Each section is supple-mentary to and not in lieu of the discussion in the SER unless otherwise noted.
Appendix A continues the chronology of the staff's actions related to the pro-cessing of the application for Braidwood Units 1 and 2.
Appendix B lists refer-ences cited in this report.* Appendix F lists principal staff members who con-tributed to this supplement.
Copies of this SER supplement are available for inspection at the NRC Publ'ic Document Room, 1717 H Street, N.W., Washington, D.C., and at the Wilmington Township Public Library, 201 South Kankakee Street, Wilmington, Illinois 60481.
The NRC Project Manager for Braidwood Station, Units 1 and 2, is Ms. Janice A.
Stevens.
Ms. Stevens may be contacted by calling (301) 492-4993 or writing:
Janice A. Stevens Office of Nuclear Reactor Regulation Project Directorate III-2 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 1.7 Summary of Outstanding Items The current status of the outstanding items listed in the SER follows:
Part A Items Status Section (1) Pump and valve operability Closed in 3.9.3.2**
Supplement 2
- Availability of all material cited is described on the inside front cover of this report.
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 4 1-1
Part A Items (Continued)
Status Section (2) Seismic and dynamic qualification of Closed in 3.10*
equipment Supplement 2 (3) Environmental qualification of electrical Closed in 3.11*
and mechanical equipment Supplement 2 (4) Containment pressure boundary components Closed in 6.2.7 Supplement 1 (5) Organizational structure Closed in 13.1, 13.4 Supplement 1 l
(6) Ersergency preparedness plans and facilities Closed in 13.3*
j Supplement 1 (7) Procedures generation package (PGP)
Closed in 13.5.2 Supplement 2 (8) Control rocm human factors review Closed in 18.2*
Supplement 4
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(9) Safety parameter display system Closed in 18.3*
Supplement 4 (10) Control room habitability Closed in 6.4 Supplement 3 Part B Items
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(1) Turbine missile evaluation Closed in 3.5.1.3 Supplement 1 (2) Improved thermal design procedures Closed in 4.4.1 Supplement 1 (3) TMI Action Item II.F.2:
Inadequate Core Closed in 4.4.7 Cooling Instrumentation Supplement 1 (4) Steam generator flow-induced vibrations Closed in 5.4.2 Supplement 1 (5) Conformance of ESF filter system to RG 1.52 Closed in 6.5.1
' Supplement 2 (6) Fire protection program Closed in 9.5.1 Supplement 3 (7) Volume reduction system Closed in 11.1, 11.4.2 Supplement 2
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 4 1-2
Status Section 1.8 Confirmatory Issues The current status of the confirmatory issues follows:
Part A Items l
(1) Applicant compliance with the Commission's Closed in 1.1, 3.1*
regulations Supplement 2 (2) Site drainage Closed in 2.4.3.3 Supplement 1 1
(3) Piping vibration test program Closed in 3.9.2.1*
l Supplement 1 (4) Preservice Inspection Program Closed in 5.2.4, 6.6*
j Supplement 2 1
'1 (5) Reactor vessel materials Closed in 5.3
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Supplement 1
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(6) Electrical distribution system voltage Closed in 8.2.4*
l verification Supplement 1 1
(7) Independence of redundant electrical safety Closed in 8.4.4 equipment Supplement 1 1
1 (8) RPM qualifications Closed in 12.5 l
Supplement 1 (9) Revision to Physical Security Plan Closed in 13.6 i
Supplement 1 (10) Control room human factors review Opened in 18.2*
Supplement 4 (11) Safety parameter display system Opened in 18.3*
Supplement 4 Part B Items (1) Inservice testing of pumps and valves Partially 3.9.6 closed in Supplement 2 (2) Steam generator tube surveillance Closed in 5.4.2.2 Supplement 1
- This section includes both site-specific-related information and duplicate-plant design features.
Braidwood SSER 4 1-3
Part B Items (Continued)
Status Section (3) Charging pump deadheading Closed in 6.3.2, 7.3.2 Supplement 1 (4) Minimum containment pressure analysis for Closed in 6.2.1.5 performance capabilities of ECCS Supplement 1 (5) Containment sump screen Closed in 6.2.2 1
Supplement 1 (6) Containment leakage testing vent and drain Closed in 6.2.6 provisions Supplement 1 (7) Confirmatory test for sump design Closed in 6.3.4.1 Supplement 1 (8) IE Bulletin 80-06 Closed in 7.3.2.2 Supplement 1 (9) Remote shutdown capability Closed in 7.4.2.2 Supplement 2 (10) TMI Action Plan Item II.D.1 Partially 3.9.3.3, closed in 5.2.2 Supplement 1 TMI Action Plan Item II.K.3.1 Closed in 7.6.2.7 Supplement 1 TMI Action Plan Item III.D.1.1 Closed in 9.3.5 Supplement 1 (11) SWS process control program Closed in 11.4.1 Supplement 2 (12) Noble gas monitor Closed in 11.5.2 Supplement 2 (13) RCP rotor seizure and shaft break Closed in 15.3.6 Supplement 1 (14) Anticipated transients without scram (ATWS)
Partially 15.6 closed in Supplement 2 (15) Evaluation of compliance with Closed in 5.2.4.4 10 cpi 50.55a(a)(3)
Supplement 2 (16) Steam generator tube failure Opened in 15.4.3 Supplement 1 Braidwood SSER 4 1-4
1.9 License Conditions The current status of the license conditions follows:
Part A Items Status Section (1) Inservice inspection program Closed in 5.2.4, 6.6*
Supplement 3 (2) Natural circulation testing Closed in 5.4.3*
Supplement 1
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(3) Response time testing Closed in 7.2.2.5*
Supplement 1 (4) Steam valve inservice inspection Closed in 10.2*
Supplement 1 (5) Implementation of secondary water chemistry Closed in 10.3.3*
monitoring and control program as proposed Supplement 1 by the Byron /Braidwood FSAR (6) TMI Item II.F.1:
Iodine / Particulate Closed in 11.5.2 Sampling Supplement 3 Part B Items (1) Masonry walls Closed in 3.8.3 Supplement 2 (2) TMI Item II.B.3 postaccident sampling Closed in 9.3.2 Supplement 1 (3) Fire Protection Program Open 9.5.1 (4) Emergency diesel engine auxiliary support Closed in 9.5.4.1 systems Supplement 3
- This section includes both site-specific related information and duplicate-plant design features.
Braidwood SSER 4 1-5
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5.2 Integrity of Reactor Coolant Pressure Boundary j
5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4.6 Evaluation of Compliance With 10 CFR 50.55a(a)(3) for Braidwood Unit 2 c
In a previous input to this section of the SER, the staff indicated that the Braidwood Preservice Inspection (PSI) Program was based on the requirements of the ASME Code Section XI, 1977 Edition with Addenda through Summer 1978.
In a letter from A. D. Miosi to H. R. Denton dated August 2, 1986, the licensee informed the NRC that a flaw was discovered in the weld between inlet nozzle F and the shell of the Braidwood Unit 2 reactor pressure vessel.
The flaw was
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discovered during an ultrasonic examination performed from the inside of the J
vessel by Combustion Engineering.
The examination indicated that the flaw was close to the outside surface and exceeded the preservice examination limits of ASME Code Section XI, 1977 Edition with Addenda through Summer 1978.
In November 1986, the licensee performed an ultrasonic examination from the outside sur-face of the vessel to locate the flaw more precisely.
This examination indi-cated that the flaw was located 1.25 inches below the outside surface and its
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size was less than the preservice examination limits of ASME Code Section XI, 3
1977 Edition with Addenda through Summer 1978.
The distance sound traveled and the amount of beam spread were greater during the ultrasonic examination from the inside surface than during the examination from the outside surface.
As the distance of sound travel and beam spread increase, the amount of flaw magnification increases.
In addition, the licensee has obtained core samples and excavated the flaws in Byron Unit 2 steani generators.
These examinations indicated that ultrasonic examination overestimated the size of the flaw.
On the basis of magnification resulting from sound travel and beam spread, and the results of previous core samplings and flaw excavations, the ultrasonic examination from the outside surface appears to give more accurate size of the flaw than the examination from the inside surface.
Since the flaw size result-ing from the ultrasonic examination from the outside surface meets the preser-vice examination limits of the ASME Code,Section XI, 1977 Edition with Addenda through Summer 1978, the reactor vessel is acceptable for service without removing the flaw.
In a letter dated December 10, 1986, the licensee proposed to perform the in-service examination of the inlet nozzle F weld from the inside surface.
If any measurable increase in the size of the defect is observed during the inservice examination, the size of the flaw will be measured by an ultrasonic examina-tion from the outside surface.
In a letter dated April 10, 1987, the licensee indicated that the ultrasonic examination from the outside surface is performed manually; the inside surface is examined automatically via remote control of the submerged ultrasonic transducer.
By ultrasonically examining and remotely handling the submerged transducer, the licensee will significantly reduce the amount of radiation to which examination personnel are exposed.
Because such exposure is to be kept as low as reasonably achievable, the staff considers the proposed inservice ultrasonic examination method acceptable for monitoring the growth of a flaw during service.
Braidwood SSER 4 5-1
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I 6 ENGINEERED SAFETY FEATURES 6.4 Control Room Habitability In a letter dated March 26, 1987, the licensee proposed operating the control room air handling unit even though its associated chiller unit was inoperable.
The licensee proposed that a temporary chilled water source be utilized during the time the air handling unit chiller was inoperable.
A cross-tie would be' made between the chilled water systems of the service building and the control By means of this cross-tie, the cooling load of the control room envelope room.
would be added to the service building's chilled water system.
The licensee indicated that this temporary chilling arrangement had been utilized during the summer of 1986 (while the control room's chilled water system was being pre-operationally tested) and provided more than adequate cooling.
The licensee is proposing this operational mode so that maintenance can be per-formed on the control room chiller units before criticality.
In SER Supple-ment 2 (SSER 2), the staff had noted the licensee's commitment to operate the control room ventilation system with one train of its emergency makeup filter system available along with its associated chiller system and air handling unit during fuel loading and reactor testing.
The licensee's proposal would modify that portion of the commitment reading, "with its associated chiller system."
The function associated with this portion of the commitment would be replaced by the cross-tie arrangement for the 14-day period.
l On the basis of its review of the licensee's proposal, the staff concludes that l
the proposed cross-tie meets General. Design Criterion (GDC) 4 of Appendix A to 10 CFR 50 as described in Standard Review Plan Section 9.4.1 of NUREG-0800 and is, therefore, an acceptable means for providing adequate temporary cooling to the control room envelope.
In SSER 2 the staff concluded that the Braidwood design meets GDC 19 even with the auxiliary building ventilation (VA) system inoperable provided:
(1) the unfiltered inleakage to the control room is limited to 25 cubic feet per minute, l
(2) emergency core cooling system (ECCS) pump leakage is limited to 1 gallon per minute, and (3) reactor power is limited to 20 percent of rated power.
By letters dated June 11, 1987, June 23, 1987, and June 26, 1987, the licensee proposed an interim operation plan for the VA system that would allow operation beyond 20 percent power until December 1, 1987.
The actual power is not to exceed a percentage of rated power (P) given by P = 97/(1+0.065L) where L is the ECCS leakage in gallons per hour.
This function represents the.
same graph proposed by the licensee in their letter dated June 11, 1987.
The staff has reviewed the proposed power limitation relationship and concludes that it is adequate to ensure compliance with both GDC 19 and 10 CFR 100 require-ments.
Therefore, the staff concludes that the power limitation relationship and Braidwood SSER 4 6-1
the program for determining ECCS leakage are acceptable.
The Unit 1 license contains a condition that the licensee adhere to the power limitations and leak rate determinations committed 4 in its letters until the VA system has been completed.
Technical Specification Section 3.7.7 has been revised to allow the VA system to be completed by December 1, 1987.
6.5 Fission Product Removal and Control System 6.5.1 Engineered Safety Feature Atmospheric Cleanup Systems In letters dated March 27, April 16, and May 11, 1987, the licensee provided
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information detailing the fact that it was unable to meet the testing criteria l
of ANSI N510-1980 for the control room recirculation charcoal adsorber.
Spe-cifically, the air flow distribution air-aerosol mixing test results were not within the acceptance criterion of 120%.
The charcoal adsorber was not de-i signed in a:cordance with the criteria of ANSI N509-1980.
However, the li-censee committed to the criteria of ANSI N510-1980 for these adsorbers.
In the letters cited above, the licensee presented the results of charcoal ad-sorber testing which indicated that the assumed efficiency of 90% for the fil-ters in the control room dose calculation is still valid despite the reduced residence time determined in the test data.
Therefore, the original conclusion that doses will be within GDC 19 limits is not affec.ed.
On this basis, the staff concludes that utilizing the recirculation adsorber f
units even though the air flow distribution and the air aerosol mixing test i
results are outside the criteria of ANSI N510-1980, is an acceptable deviation from the Standard Review Plan since doses are within the requirements of GDC 19.
The sta'f, therefore, considers this issue resolved.
Braidwood SSER 4 6-2 i
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9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection Program 9.5.1.5 Fire Protection for Specific Plant Areas In its letters dated June 10 and June 18, 1987. the, licensee indicaud that the installation of penetration seals in fire walls and fire-rated floot eiling assemblies and the installation of fireproofing for steel structura; elements l
will not be completed oefore exceeding 5% of rated power as stated in SER Sup-l plement 2.
The licensee also stated in these letters that roving fire watch patrols have already been instituted for these affected areas and will remain in effect as required by the Station Administrative Procedures until all of the installation work has been completed by August 31, 1987.
On the basis of this compensatory measure, the staff concludes that an adequate level of fire safety will be provided pending completion of the required modifications.
Therefore, this matter is considered resolved.
i 9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System 9.5.4.2 Emergency Diesel Engine Fuel Oil Storage and Transfer System Design
- The SER stated that in order to minimize stirring up sediment when diesel fuel oil is added to the Unit 1 fuel oil storage tanks, the twin diesel fuel oil storage tanks for each diesel generator vill be replenished by refilling one I
tank while the other tank provides fuel oil to the diesel. The fuel oil in the refilled tank will then be allowed to settle for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before the other tank is refilled or before fuel oil is suctioned from the refilled tank.
In a letter dated February 25, 1987, the licensee stated that it intended to l
l perform a 100-hour reliability run on both diesel generators of Braidwood Unit 1.
To perform this test without violating the Technical Specification requirement of maintaining a minimum 44,000 gallons of fuel oil in an operating plant, the I
I licensee would have to continually refill the fuel oil storage tanks to main-
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tain that minimum.
Upon replenishing the tanks, the licensee would have to follow the FSAR/SER commitment stated above for refilling the fuel oil storage tank.
However, the settling time restriction could not be satisfied for this test run.
In the letter dated February 25, 1987, the licensee proposed that as long as the fuel oil levels in each tank are maintained above the 50% level, the sedi-ment in the tank would be stirred minimally and waiting time for settling can be eliminated.
Should the level fall below 50% in a tank, the original FSAR/SER
- Title of Section 9.5.4.2 has been changed from SER to more accurately reflect the contents of this section.
Braidwood SSER 4 9-1 l
commitment would' apply.
The staff has evaluated this proposal'for other plants with similar tank designs and has found that it is an acceptable way of meeting Position C.2.g of Regulatory Guide 1.137 on minimizing stirring up' sediment in i
fuel oil storage tanks while ensuring that'a minimum of 44,000 gallons of fuel oil is available.
Therefore, the staff finds the proposed change to the plant I
cperating procedures /FSAR commitment acceptable.
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Braidwood SSER 4 9-2 l
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18 HUMAN FACTORS ENGINEERING 18.2 Main Control Room and Remote Shutdown Panel In its letter dated December 1, 1986, the licensee submitted the Detailed Con-trol Room Design Review (DCRDR) Summary Report.
This summary report included the human engineering discrepancies (HEDs) applicable to the Braidwood site-specific panels and instrumentation, and the survey findings along with proposed i
corrective actions and implementation schedules.
The staff is satisfied with 1
all proposed control room improvements and the schedule for their implementation.
Therefore, Outstanding Item A(8) is closed, j
On the basis of a preimplementation site audit conducted on March 10-11, 1987, the staff concluded that all requirements of Supplement 1 to NUREG-0737 had been satisfactorily completed except for:
- 1) the evaluation of the long term engineering solution to the problem of radio transmitters that can inadvertently activate safeguard systems; 2) the evaluation of the lack of lamp test capabil-ity; 3) the determination of a long-term engineering solution to the indicator light bulb burnout problem; 4) the documentation of a detailed color versus use I
matrix for control room color coding and a commitment to make color coding in all control room contexts consistent with the green board concept; and 5) the documentation of a clarification of the licensee's response to HED No. 0165 regarding l
computer printer speed, in order to provide assurance that data will not be lost during a major event.
In its letter dated June 23, 1987, the licensee committed to submitting pro-l posed resolutions to these five items by August 31, 1987.
The staff finds I
this schedule acceptable.
This is Confirmatory Issue A(10).
With the control room design improvements that have already been completed, the staff has deter-i mined that the potential for operator error leading to serious consequences as a result of human factors considerations in the control room is sufficiently i
low to permit safe operation of Unit 1.
18.3 Safety Parameter Display System l
In its letter dated June 23, 1987, the licensee stated that the identification of the wide-range and narrow-range iconic displays had been corrected for c
Braidwood Unit 1 in order to satisfy the requirements of Supplement 1 to NUREG-0737.
Therefore, Outstanding Item A(9) is closed.
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1 On the basis of a preimplementation site audit conducted on March 10-11, 1987, j
the staff concluded that all requirements of Supplement 1 to NUREG-0737 had 1
been satisfactorily completed except for several operational problems with the SPDS. On the wide-range iconic screen, four of the eight parameters varied from the expected value enough to detract from the display.
These were as follows.
1 (1) The reactor vessel level indication system (RVLIS) was reading properly but the value spiked to the alarm level periodically.
Braidwood SSER 4 18-1
(2) The radiation level remained at the alarm setpoint.
l (3) The steam generator level was reading far enough from the setpoint to be distracting.
(4) The containment pressure was inaicating far enough from the setpoint to skew the iconic display.
On the narrow-range screen, the following three readings were off the setpoints:
(1) high radiation alarm (2) containment temperature (3) steam generator level In its letter dated June 23, 1987, the licensee committed to submitting pro-posed resolutions to these items by August 31, 1987.
The staff finds this schedule acceptable.
In its letter dated October 10, 1986, the licensee committed to submitting the verification / validation report by August 1, 1987.
Confirmation that the SPDS operational problems have been corrected and sub-mittal of the verification / validation report constitute Confirmatory Issue A(11).
1 Braidwood SSER 4 18-2
APPENDIX A CONTIN 0ATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL SAFETY REVIEW 0F BRAIDWOOD STATION, UNITS 1 AND 2 l
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April 10, 1987 Letter from licensee concerning reactor vessel nozzle' analysis.
1 April 28, 1987 Letter from licensee concerning Inservice Inspection (ISI)
J Program.
1 April 28, 1987 Letter from licensee concerning seismic Category I manhole Covers.
April 29, 1987 Letter from Isham, Lincoln & Beale concerning inadvertent l
disclosure of materials subject to the Atomic Safety and Licensing Board protective order.
April 29, 1987 Letter from licensee concerning containment leak chase channels.
May 6, 1987 Filing from the Atomic Safety and Licensing Board Panel concerning motion to admit late-filed contention on finan-cial qualifications.
May 6, 1987 Letter from licensee concerning removal of control room chlorine detectors.
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May 13, 1987 Filing from the Atomic Safety and Licensing Board concerning Partial Initial Decision on Emergency Planning Issues.
1 May 15, 1987 Filing from Atomic Safety and Licensing Appeal Board concern-1 ing response to interveners' motion seeking to reopen the record for the admission of a new contention.
May 18, 1987 Letter from licensee concerning Atomic Safety and Licensing Board's decision on emergency planning.
May 18, 1987 Filing concerning reconstitution of Atomic Safety and Licensing Appeal Board.
May 18, 1987 Memorandum from Atomic Safety and Licensing Appeal Board concerning reconstitution of Board.
May 19, 19E,7 Filing from the Atomic Safety and Licensing Board concerning concluding Partial Initial Decision (0perating License).
Braidwood SSER 4 1
Appendix A
l-May 19, 1987 Filing from the Atomic Safety and Licensing Board concerning corrected pages to Minority Opinion.
May 26, 1987 Filing from the Atomic Safety and Licensing Board concerning NRC Staff Response to Motion to Admit Late-Filed Contention on Financial Qualifications.
May 28, 1987 Letter from licensee concerning application for amendment.
May 28, 1987 Transcript concerning Immediate Effectiveness Issues.
May 29, 1987 Letter from Isham, Lincoln & Seale to the Atomic SaTety and Licensing Board concerning the withdrawal of licensee's jurisdictional objection to intervenor's motion tn admit late-filed contention.
June 1, 1987 Filing from the Atomic Safety Jurisdictional and Licensing Board concerning Notice of Appeal.
June 1, 1987 Letter from licensee concerning Emergency Response Facilities.
June 3, 1987 Letter from Isham, Lincoln & Beale to the Atomic Safety and Licensing Board concerning forwarding the May 28, 1987 appli-cation for amendment June 3, 1987 Letter to licensee concerning request for information concerning offsite medical services for Units 1 and 2.
June 9, 1987 Filing from the Atomic Safety and Licensing Board concerning emergency planning order.
June 1-0, 1987 Filing from the Atomic Safety and Licensing Board concerning notice of reconstitution of board.
June 10, 1987 Filing from the Atomic Safety and Licensing Board concerning Memorandum and crder denying interveners' motion to admit late-filed contentions on financial qualifications.
June 11, 1987 Letter from licensee concerning Interim Operation of VA System.
June 12, 1987 Transcript of telephone conference in Washington, D.C.
June 15, 1987 Filing from the Atomic Safety and Licensing Board concerning memorandum on licensing board jurisdiction.
June 23, 1987 Letter from licensee concerning detailed control room design review and safety parameter display system.
June 23, 1987 Letter from licensee concerning Interim Operation of VA System.
June 26, 1987 Letter from licensee concerning Interim Operation of VA l
System.
Braidwood SSER 4 2
Appendix A
APPENDIX B BIBLIOGRAPHY American National Standards Institute /American Society of Mechanical Engineers, N509, " Air Cleaning Units and Components, Nuclear Power Plant," 1980.
--, N510, " Testing of Nuclear Air-Cleaning Systems," 1980.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, I
Section XI, " Rules for Inspection of Nuclear Power Plant Components," 1977 Edition with Addende through Summer 1978.
4 i
U.S. Nuclea" Regulatory Commission, NUREG-0737, Supplement 1, " Clarification of TMI Action Plan Requirements," January 1983.
l l
Braidwood SSER 4 1
Appendix B l
i
l APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS NRC STAFF Name Title Review Branch
- John W. Craig Branch Chief Plant Systems Branch, DEST Richard J. Eckenrode Human Factors Engineer Human Factors Assessment Branch, DLPQ Barry J. Elliot Materials Engineer Materials Engineering Branch, DEST Robert J. Giardina Mechanical Engineer Plant Systems Branch, DEST John J. Hayes, Jr.
Nuclear Engineer Project Directorate II-1, DRP-I/II Dennis J. Kubicki Fire Protection Engineer Plant Systems Branch, DEST Calvin W. Moon Senior Reactor Engineer Technical Specifications, DOEA Rayleona F. Sanders Technical Editor Policy & Publications Management, DPS Jerry J. Swift Health Physicist Radiation Protection Branch, DREP Catherine S. Vogan Licensing Assistant Project Directorate I-1, DRP-I/II Jared S. Wermiel Section leader Plant Systems Branch, DEST
- Reflects reorganization since SER was issued.
Braidwood SSER 4 1
Appendix F
y p
NR F0hM 336 U E. NUCLE 13 ? t1ULAT&AY COMMISSION i WE OWT NUMS&M #Assga.dk IlO add F.i ifA, st.eyl NUREG-1002 f ik'"2E' BIBLIOGRAPHIC DATA SHEET Supplementflo. 4 utmstr.UefioN o i.Evias.
anca sNa su.Taa s a a n euANK y
- f Safety Ev uation Report related to the 7
operation f Braidwood Station, Units 1 and 2 f.4 DATE REPOAT COM*tETED MONTH TEAM l
dbne 1987 U, o.
6 OATE REPORT ISSUED
.c MONTM VSAR A
July 1987 i
I 6E A8OnMING QRGAN62ATION NAME A MAIL,NG A00mgss gieve le c.ari
,8 PROJECTIT ASK/ WORK UNIT NUMetN Division of Reac or Projects - III, IV, V and /
Special Projec s ff e n~ oa ca A~T NuMua i
Office of Nuclear eactor Regulation g
i U. S. Nuclear Comm sion f
Washington, D. C. 2 55 r
to
-ON.o.mo o.o ANa ATio,. NAM. ANo MA <,N ooaus <,
- s. c,
ii. TvPE o a:PoaT p
n' Same as 7. above
[
T c nica1 t
,,g, o9 g9 n,,,,,,,,,,,,,,,,
- p November 1983 -
June 1987 12 $UPPLEMENT ARY NOTts
-1 4p' Docket Nos. STN 50-456 and.STN 5.01457
,, Aes,e Acum t
p In November 1983, thestaffo'NheNuclearRegulatoryCommissionissued its Afety Evaluation Report (m REG-1002) regarding the application filed by the Commonwealth Edison Cfm' ny, as applicant and owner, for a license to operate Braidwood StatioJ)', U -its 1 and 2 (Docket Nos. 50-456 and
)
50-457).
The first supplement t
NUREG-1002 was issued in September 1986; thesecondsupplementtogUREG-10 supplement to NUREG-1002pias issu,2 was issued in October 1986; the third in May 1987.
This fourth supplement to NUREG-1002 reports the status o,certain items that remained unresolved at the time gupplement 3.as published.
The facility is locatedinReedTownspp,WillCounti Illinois.
~
fy 1
.. ooCuMiNT AN At..... o vyoi, ooc.,,1o,
STATEMENT j
Unlimited 16 BECURITY CLAS$1FICATION 6 LOfMTfFIEA$10 PEN EN TERMS llNCIASSIFIEC 17 NUM9tH of P AGt5 IS PRICE
- U,1. Covinhal[hi PAINithG OUIC[i1987.181-682:63153 l
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