ML19257A002
| ML19257A002 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/14/1979 |
| From: | Green W ENERGY, INC. |
| To: | |
| Shared Package | |
| ML19256G557 | List: |
| References | |
| CPR-001-01, CPR-1-1, NUDOCS 7912310473 | |
| Download: ML19257A002 (47) | |
Text
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ENERGY INCORPORATED p
DESCRIPTION TEST SPECIFICATION FOR PRESSURIZER RELIEF VALVE POSITI0tl INDICATION SYSTEM AT BIG ROCK POINT NUCLEAR PLANT.
DATE PREPARED BY M 0A Aa %
18-//4/77 CONTRACT NO. 13090-000
/l//4/7 REVIEWED BY
'e1 '-
RELEASE DATE 12/14/79 w
IFdf QUALITY ASSURANCE A /
/
PREPARED FOR Consumers PROJECT MANAGER A /MM A -v
/2,/"e/'9 Point Plant
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7 G.W.0. 6971 REVISIONS 1661 147 REVISION NO.
T-SPEC NO.
CPR-001 0
l PAGE _ OF 14 JOB CODE-S/N 791231OV-73 -2
TABLE OF C0flTEtlTS PAGE 1.0 PURP0SE............................................................
3 2.0 R EF ER E NC E 00C UME NTS................................................ 4 2.1 0rawings......................................................
4 2.2 S pe c i fi c a t i o n s................................................ 4 3.0 EQUIPMENT LIST.....................................................
5 4.0 P R 0C E D UR E.......................................................... 6 4.1 Transducer Cables Grounding and Continuity....................
6 4.2 Cabling From Charge Ampl ifier To Control Room................. 8 4.3 AC P o wer C i rc u i t............................................. 12 4.4 Lab el i ng, Ma rki ng, Reco nnectio n.............................. 13 1661 148 REVISION NO.
T -SPEC NO.
CPR-001 0
JOB CODE-S/N PAGE LOF la 73-2 FORM E-034 REV.0,lo/79
1.0 PURPOSE This test procedure will be used to verify correct installation of field wiring and cabling. Continuity of circuits and correct grounding will be tested.
The tests in this specification are to be done after installation of cable and wiring is complete fra the transducer area to the control room.
These tests shall be completed prior to installing transducers or connecting power to the monitor panel. These tests are preparatory to perfoming the operational testing and calibration.
1661 149 REVISION NO.
T -SPEC NO.
cPR-001 0
JOB CODE-S/N PAGE 3 op 14 73-2
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t 2.0 REFERENCE DOCUMENTS 2.1 Orawings EI Drawing 9430 Schematic Wiring Diagram, Steam Drum SRV, Position Indication Systen EI Drwing 9425 Steam Drum, SRV Position Indication System, Installation Details EI Drawing 9426 Charge Amplifier Junction Box Detail, Steam Drum SRV, Position Indicator Systen 2.2 Soecifications EI Construction Specification C-Spec-CPR-001 El Construction Specification C-Spec-CPR-002 1661 150 REVISION NO.
T - SPEC NO.
CPR -001 0
JOB CODE-S/N PAGE i OF li 73-2
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3.0 EQUIPMENT REQUIRED The following equipment is required to perform the test and ekeckout.
(1) Ohmmeter, Simpson Model 260 or approved equal.
(2) Clip leads.
CAUTION: Do NOT use Megger.
1661 151 REVISION No.
T - SPEC NO.
CPR-001 0
JOB CODE-S/N PAGE l_OF14 73-2
$$.5,ioM
4.0 PROCEDURE 4.1 Transducer Cables Grounding and Continuity This step will test for continuity and for grounds in the cables connecting the accelerometers to the charge amplifiers.
(1) Remove the connector from transducer, FE 5000, if it has been installed. Remove the other end of the cable from charge amplifier, CA 5000; that is, the softline coaxial cable on the input of the charge amplifier.
(2) Using an ohmmeter, measure the resistance to ground of the cable center conductor and of the cable shield, and record belcw.
(3) Measure and record resistance from center conductor to shield.
(4) Measure and record continuity of the cable center conductor and of the cable shield from the charge amplifier to the accelerometer connection.
Using a clip lead short the center conductor to the shield at one end and measure resistance between conductor and shield at the other end.
(5) Replace the connector on the charge amplifier and if the transducer is installed replace the cable connector, safety wire, if provided for, and seal connector with Dow Corning RTV Type 96-081.
(6) Repeat the above procedure on the remaining five cables.
1661 152 REVISION NO.
T - SPEC NO.
Cea-001 0
JOB CODE-S/N PAGE 1 0F 14 73-2
$3.5fo'P//
Measured Estimated Cable No. 6813-ZO2-ZO3/1 Resistance Resistance (2) Center conductor to ground
> 10 meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Center conductor in series with shield
< 30 ohms Cable No. 6813-ZO2-ZO3/2 (2) Center conductor to ground
> 10 meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Center conductor in series with shield
< 30 ohms Cable No. 6813-ZO2-ZO3/3 (2) Center conductor to ground
> 10 meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Center conductor in series with shield
< 30 ohms Cable No. 6813-ZO2-ZO3/4 (2) Center conductor to ground
> 10 Meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Center conductor in series with shield
< 30 ohms 1661 153 REVISION NO.
T - SPEC NO.
CPR-001 0
JOB CODE-S/N PAGE 7 0F 14 73 -2
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Cable No. 6813-202-Z03 /5 (2) Center conductor to ground
> 10 meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Center conductor in series with shield
< 30 ohms Cable No. 6813-ZO2-Z03/6 (2) Center conductor to ground
> 10 meg (2) Shield to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Ces;ter conductor in series with shield
< 30 ohms Step 4.1 tests certified complete and results satisfactory:
Test Technician Date 4.2 Cabling From Charge Amplifier to Control Room This step will test for continuity and for grounds in the cables connecting the charge amplifiers to the signal conditioners in the Control Room.
(1) Remove the twisted shielded pair from the output of charge amplifier CA-5000. Remove the coaxial connector from the input of signal conditioner FI 5000. Remove the ground wire from C-3, 6, 9, and 12 and from D-3 and 6.
(2) Measure the resistance to ground of the center conductor of the coaxial cable and record below.
1661 154 REVISION NO.
T -SPEC NO.
CPR-001 0
8 PAGE
,,,,,op 14 JOB CODE-S/N 73-2
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(3) Measure and record the resistance from center conductor to shield.
(4) Measure and record the resistance to ground of the coaxial connector shield.
(5) Measure and record the resistance between the twisted pair shield, C-3, and ground.
(6) Measure and record the resistance from the coaxial outer conductor to C-3.
(7) Measure and record the resistance from the coaxial center co1ductor to C-3.
(8) At the charge amplifier J-box, clip the wires C, S and SH together. Measure and record the resistance between the center conductor and shield on the coaxial connector at the signal
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conditioner input.
(9) Measure and record the resistance between the center conductor, on the coaxial connector at the signal conditioner, and teminal C-3.
(10) Replace the ground wire on C-3.
(11) Measure and record the resistance from signal conditioner chassia to ground.
(12) Replace the connector on the signal conditioner.
Replace the wires on the charge amplifier.
(13) Repeat the above procedure on the other five channels.
1661 155 REVISION NO.
T - SPEC NO.
CPR-001 3
JOB CODE-S/N PAGE 9 0F 14 73-2
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Measured Estimated Cable flo. 6813-Z01-ZO2/1 Resistance Resistance (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, C-3
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground teminal, C-3 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 ohm Cable flo. 6813-Z01-ZO2/2 (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, C-6
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground teminal, C-6 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 chm 1661 156 REVISION NO.
T -SPEC NO.
CPR-001 O
JOB CODE-S/N PAGE l.20F 14 73-2
$E$.5,io8/
Cable No. 6813-Z01-ZO2/3 (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, C-9
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground teminal, C-9 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 chm Cable No. 6813-Z01-Z02/4 (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, C-12
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground teminal, C-12 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 ohm 1661 157 REVISION NO.
T -SPEC NO.
CPR-001 0
JOB CODE-S/N PAGE.ll_OF 14 73-2
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Cable No. 6813-Z01-ZO2/5 (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, D-3
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground terminal, D-3 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 chm Cable No. 6813-201-Z02 /6 (2) Center conductor to ground
> 10 meg (3) Center conductor to shield
> 10 meg (4) Coaxial shield to ground
> 10 meg (5) Twisted pair shield to ground
> 10 meg (6) Outer conductor to twisted pair shield, D-6
> 10 meg (7) Center conductor to twisted
> 10 meg pair shield (8) Center conductor and outer conductor in series
< 8 ohms (9) Ground terminal, 0-6 to center conductor
< 8 ohms (10) Signal conditioner chassis to ground
< 1 cha 1661 199 REVISION NO.
T -SPEC NO.
CPR-001 0
JOB CODE-S/N PAGE.12,,op 14 73-2
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Step 4.2 tests certified complete and results satisfactory:
Test Technician Date 4.3 AC Power Circuit This step will test the AC Power circuit to the Audio Monitor and signal conditioners for grounds and continuity.
(1) Verify that, the 5 A. fuses in the power circuit to teminal block CC are removed.
(2) Measure resistance from CC-81 and CC-84 to ground, and record b el ow.
(3) Measure and record resistance from CC-81 to CC-84. A reading less than about 10 ohms indicates a short, greater than 30 ohms indicates an open circuit.
(4) Measure and record resistance from fuse teminal 2 to CC-81.
(5) Measure and record resistance from fuse teminal 1 to CC-84.
(6) Measure and record voltage between the fuse teminals on the hot side.
Power Circuit Measured Estimated (2) CC-81 to ground
> 10 meg (2) CC-84 to ground
> 10 meg (3) CC-81 to CC-84
> 10 ohms < 30 ohms (4) Teminal 2 to CC-81
< 1 ohm (5) Teminal 1 to CC-84
< 1 ohm (6) Voltage at hot terminals 120 t 10 volts AC 1661 159 REVISION NO.
T -SPEC NO.
CPa-001 0
JOB CODE-S/N PAGE.L_OF 14 73-2 EM Ei2
If the measurements are satisfactory insert fuses.
Step 4.3 tests certified complete and results satisfactory:
Test Technician Date 4.4 Labeling, Marking, and Reconnection (1) Verify that all cables are labeled at each end with the cable number.
(2) Verify that all connectors are labeled with the connector number.
(3) Verify that all wires teminating on a teminal strip are labeled with the teminal number.
(4) Verify that all transducers, charge amplifiers and signal conditioners are marked with identifying numbers.
(5) Reconnect all connectors, replace all wires on teminals to leave the system operational.
(6) Mark all prints with the "as-built" conditions. These marked drawing will be used to prepare "dr6 wings for record".
1661 160 REVISION NO.
T -SPEC NO.
cPR-001 0
JOB CODE-S/N PAGE.L 0F 14 73-2
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APPENDIX C Design Review of Plant Shielding 1661 167
1 I.
Description of Facility Big Rock Point is a 240 MWt boiling water reactor located near Charlevoix, Michigan. The reactor is contained within a 130-foot diameter steel hortonsphere. The hortonsphere is constructed of 3/4-inch steel plate. The plant was constructed in the early 1960s and began operation in 1962.
Big Rock Point, as is the case for other reactors of similar vintage, has no shield outside the containment building.
Big Rock Point is included in NRC's ongoing Systematic Evaluation Program (SEP). The purpose of SEP is to evaluate certain older operating reactors with respect to current NRC licensing criteria.
Differences from current criteria are evaluated to determine equivalence / acceptability. Differences not determined equivalent will be considered during an " integrated assessment" near the end of the SEP to determine wbich, if any, must be eliminated by plant modification /backfitting of current criteria.
Modifications to the plant resulting from SEP, if any, would therefore generally be accomplished following the end of the program. SEP review of Big Rock Point is currently scheduled to end in May 1982.
II.
Review of Existing Facility NUREG 0578 requires an assumption of 100% core inventory of radioactive noble gases and 25% core inventory of radioactive halogens dispersed in the containment atmosphere. The direct radiation from such an assumed source far exceeds radiation levels from any other source.
In fact, plant design is such that most systems which might contain radioactivity after an accident are within containment, thereby decreasing the number of potential sources of radiation which must be considered.
Calculations for the early post-accident stages, using the above assumptions, yield estimated radiation levels for direct radiation from the containment atmosphere well in excess of 10 R/h.
These doses decrease with time as short-lived radionuclides decay. During the initial hours after a postulated accident which results in the release of radioactivity at the assumed levels, however, movement around the site would be essentially precluded by " containment shine."
Certain areas on site are sufficiently shielded that they would remain tenable following such a postulated accident. These areas p-incipally are the control room, the shif t supe visor's office adjacent to the control room (currently designated the interim Technical Support Center), and the area beneath the control room (currently designated the Operational fupport Center). Radiation levels in these areas, for example, would be approximately 15 mR/h one hour after a postulated accident which released the specified fractions of core inventory.
Actions which must be taken after an accident have been reviewed to determine if modifications are required in order to limit personnel radiation exposures to 25 Rem in the 30 days following an accident. This review considered actions which necessarily must be performed (such as manual switchover from 1661 168
2 injection to recirculation mode of core cooling) and actions which might be required ir. the event of certain failures. The modifications considered necessary to ensure the desired limitation of personnel exposures using the assumptions specified by NUREG 0578 are:
Backup Emergency Dtesel Generator The backup emergency diesel generator is currently mounted in a semitrailer located on site.
If needed, it is moved to the required location and connected to the plant emergency power bus.
Evaluation of this operation under the above specified post-accident conditions indicates personnel radiation exposures above 25 Rem could be received performing these actions.
The backup emergency diesel generator will be relocated to an area near the primary diesel generator. Modifications will be made to enable the backup to be placed in service remotely from the control room.
Backup Cooling Water Supply to the Core Spray Heat Exchanger Long-term core cooling after an accident is accomplished by recirculating water from the containment building sump through a heat exchanger and back to the core via the core sprays. Cooling water for the heat exchanger is supplied by the fire water system. Plant Technical Specifications require a backup hose to provide cooling water to the core spray heat exchanger in the event of a failare of the buried fire main. Evaluation of this system has shown that personnel exposures above 25 Rem could be received installing the backup hose under the above specified post-accident conditions.
A permanent core spray heat exchanger water supply line, separate from the current buried fire main and fulfilling the function of the backup hose, will be installed. This line will be capable of being placed in service remotely from the control room.
Emergency Diesel Generator Fuel Supply The emergency diesel generator currently has a fuel supply capable of sustaining operation at full load for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Evaluation of the refueling operation under the above specified post-accident conditions indicates personnel radiation exposures above 25 Rem could be received if refueling occurred 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident.
A new fuel supply tank will be installed. The new tank will provide a seven-day fuel supply. Refueling after seven days under the above specified post-accident conditions should not result in personnel radiation exposures above 25 Rem.
Preliminar/ engineering for these modifications has begun. These modifications will be installed prior to January 1, 1981.
let!1 59
3 Certain operator actions have been identified which might be specified by procedure (eg, response to certain plant alarms) but which would not be essential in a post-accident situation.
These procedures cannot simply be changed, however, because many of these na ae procedures govern operation during nonaccident conditions. Precautiocary statements addressing situations following receipt of a containment radiation monitor alarm will be incorporated into plant procedures, as necessary. These statements will caution the operator to evaluate the need to perform actions outside the control room against the radiation exposure which may be received. These precautions and a procedure for performing the required evaluation will be in place prior to start-up from the outage scheduled to begin in late December 1979.
The review performed also identified a need to control plant ingress and egress after an accident. To this end, plant procedures will be revised to specify that post-accident shift changes and entry to the site by off-site personnel not be permitted unless radiation levels are acceptable for transitting to/from the control room / Technical Support Center / Operational Support Center. Off-site personnel who would otherwise be recalled to th<.
Tehenical Support Center will, under these conditions, be directed to the near-site support center. Our evaluation indicates that radiation levels will decrease to a level permitting plant ingress / egress approximately 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> af ter a postulated accident which releases the core inventory fractions specified in NUREG 0578.
III. Long-Term Modifications Additional shielding will be installed, in the long term, to enable personnel radiation exposures to be controlled to 10 CFR 20 levels under the specified conditions. For reasons discussed at length below, this installation will be incorporated into modifications which may result from the SEP.
Actual installation of additional shielding will, therefore, take place after the end of SEP.
Consumars Power Company has engaged a contractor, Catalytic Inc, to perform an avaluation and conceptual design of different options of meeting the requirements of NUREG 0578. This conceptual design has been accomplished now, despite the intention to perform actual installation of additional shielding af ter the end of SEP, in order to permit engineering efforts to begin shortly.
These engineering efforts must proceed in parallel with SEP to generate the information required to be integrated into SEP topic review.
(See V below.)
The conceptual design review is complete.
Catalytic has recommended three options for consideration. These are (1) installing local shielding for specific plant areas, (2) erecting a concrete shield building around the existing containment, and (3) a combination of these methods. Each of these options is discussed briefly below.
The conceptual design work performed by Catalytic was based on the following assumptions:
1661 170
4 1.
Core radionuclide release fractions specified in NUREG 0578.
2.
Instantaneous mixing of released radionuclides in the containment atmosphere (at time t = 0).
3.
Conservatively established criteria for post-shielding radiation levels in areas of concern.
The conservative nature of these assumptions resulted in conceptual designs which are considered to be an upper bound on the modifications which will actually be required. Catalytic's analyses considered only the radionuclides mixed in the containment atmosphere since containment water level remains significantly below grade level. Shielding afforded by structures inside the containment was not considered.
The option of local shielding only was designated by Catalytic as Option A.
A conceptual drawing of the shielding which might be required for this option (SK-1300-1, Rev P) is included herein. This option includes addition of concrete shielding to existing structures in areas considered essential for post-accident availability.
Concrete-enclosed walkways would be provided to facilitate plant ingress / egress and movement between essential areas. A staging area, with shadow shield wall, would be established at the end of the site access road for use by personnel responding from off site.
A shielded walkway would extend from the staging area to the security building and, thence, into the protected area.
Shield wall thicknesses indicated on the conceptual drawing are likely upper bounds due to the conservative assumptions discussed above.
Catalytic's Option F consists of enclosing the existing containment within a shielding building of sufficient thickness to preclude the need for other shielding. The conceptual estimate (conservative) of the amount of shielding needed is a 56-inch thick concrete silo with a 30-inch thick concrete roof.
This option is shown on enclosed Drawing SK-1300-12, Rev P.
Option B is the combination of A and F.
In this option, the proposed shielding building is 36 inches thick and local shielding is still required for some of the areas which would be shielded under Option A.
This option is depicted on enclosed Drawing SK-1300-6, Rev P.
Again, chield thicknesses shown are probably above that which detailed design would show to be needed.
Preliminary cost estimates for the above-described options range from $35 to
$40 million. Preliminary schedules for design, engineering, procurement, and construction indicate a total project duration of 22-24 months including extended plant outages of five months cr more.
IV.
Qualification of Critical EquipmeSt Requirement 2.1.6.b of NUREG 0578 also specifies that equipment which must function after an accident be identified and reviewed to assure it can withstand post-accident radiation exposures it might receive.
Necessary equipment has been determined, based upon information generated during review of various SEP topics. A tabulation of this equipment is attached. The 1661 171
5 tabulation also indicates the radiation dose to which each item of equipment would be subjected under the accident assumptions specified in NUREG 0578 (assuming existing plaut design).
The radiation dose to which equipment outside containment might be subjected is likely to change with installation of additional shielding as discussed above. Review of this equipment to assure capability to withstand the expected dose will be accomplished soon after the final shield design is completed. The remaining equipment will be reviewed and modified / replaced, as necessary, prior to January 1, 1981.
It should be noted that control room equipment is not included in the attached tabulation since the existing control room shield is sufficient to eliminate concern about radiation doses for equipment it protects.
V.
Need To Integrate With SEP The massive amounts of shielding which would be added by any of the options discussed above would significantly impact several SEP topics.
Chief among them is seismic design. Due to size and proximity to safety equipment, any added shield would have to meet seismic design criteria. The SEP is intended to generate new seismic criteria for the Big Rock Point site to replace that used in initial plant design. New seismic criteria are expected to be available during the first half of 1980.
In parallel with development of these criteria, existing plant structures are being modeled for seismic analyses. Any shielding added to these structures would require modeling and seismic analyses before it can be determined that the existing structures, plus proposed modifications, are acceptable. The iterative procedure of design and analysis can begin only after the new site seismic criteria have been established.
It is, therefore, not possible to design additional shielding on the schedule specified in NUREG 0578.
The high cost of the required shielding modification necessitates particular care in its design.
In particular, criteria which might be generated by other SEP topics (eg, tornado missiles) should be incorporated in its design. It is, also, possible that SEP may identify the need for new safety equipment at Big Rock Point (eg, a dedicated or " bunkered" safe shutdown system). The nature of the shielding design options available is such that it would be likely to seriously interfere with the installation of any additional safety systems unless the designs were integrated.
Optimization of plant operability and the human engineering aspects of system design could best be achieved by integrating the design of various possible modifications. This optimization is an end goal of SEP and forms a large portion of the basis for the SEP design which ends with an " integrated assessment." Long-term safety would best be served by a detailed integration of shelding design and any modifications which might result from SEP.
1661 172
6 VI.
Justification for Continued Operation The review described in II above has determined that the radiation exposures to plant personnel can be limited to below life-threatening levels even in the unlikely event of an accident of the severity postulated in NUREG 0578. The high radiation levels which would be produced by such an accident would be localized and affect the plant site only.
No appreciable increase in radiation exposure to any off-site individual would be expected. The effect on public health and safety of such an accident occurring before erection of any additional shielding is, therefore, small. The probability of such an accident occurring in the brief period before the end of SEP is vanishingly small. These facts, combined with consideration of the many actions being taken at Big Rock Point in response to other NUREG 0578 requirements, clearly indicate that continued plant operation pending integration of shielding design with SEP is justified.
1661 173
I EQUIPMENT TO BE CONSIDERED IN POST-ACCIDENT RADIATION DOSE EVALUATION Integrated Dose Over Eigu i pmen t Function Time Needed Time Needed (Rads) 5 6
1.
CV-4094 Sphere Exhaust Valve 30 Days 4 x 10 to 1.1 x 10 CV-4095 CV-4096 Sphere Vent Valve CV-4097 5
6 2.
SV-9151 Air to CV-4096 30 Day, 4 x 10 to 1.1 x 10 SV-9152 Air to CV-4097 SV-9;53 Air to CV-4094 SV-9154 Air to CV-4095 SV-9155 Isolation of Vent Probe SV-9156 5
6 3.
C-26 Panel Air Shed Panel 30 Days 4 x 10 to 1.I x 10 5
6 4.
PCV-4514 Nitrogen Supply to 30 Days 4 x 10 to 1.1 x 10 Vent Valves.
(Backup Heans to Open Vent Valves During Vacuum Relief Phase.)
4 6
5.
POIS 7814 Strainer Differential Pressure 30 Days 2x 10 to 1.1 x 10 POIS 7815 POIS 7816 6.
TIA Personnel Lock Remain Scaled 1.1 x 106 (Integrated TIB E<luipment Lock Indefinitely over 30 Days)
TIC Escape Lock Associated Equalizing Valves 5
6 7.
PS-664 Enclosure Pressure Short Period. Thereafter 4x 10 to 1.1 x 10 PS-665 Remain Scaled Indefinitely (Integrated over
~
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PS-667 Ch
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2 Integrated Dose Over Equipment Function Time Needed Time Needed (Rads) 8.
PS-636 NO-7068 Timer Start Containment Spray 4 x 105 to j,j x 106 PS-637 MO-7068 Timer Start Delay Needed Only 20 Minutes.
(Integrated Over PS-70644 NO-7064 Timer Start Thereaf ter Remain Sealed 30 Days)
PS-7064B MO-7064 Timer Start Inde fi ni tely.
9.
HO-7067 Turbine Bypass isolation 30 Days 1x 105 to 4 x 105 CV-4106 Bypass Warmup Control Valves Must SV-4916 Air to CV-4106 Stay Closed CV-4104 Steam to Seals and Air Ejector SV-4899 Air to CV-4104 CV-4200 Turbine Stop Valve CV-4117 Reactor and Fuel Pit 1) rain Isolation SV-4922 Air to CV-4117 CV-4102 Clean Enclosure Sump Isolation SV-4895 Air to CV-4102 CV-4103 Dirty Enclosure Sump Isolation SV-4896 Air to CV-4103 CV-4105 Demin Water Isolation SV-4897 Air to CV-4105 10.
Electric Fire Pump ECCS 30 Days 1 x 105 Diesel Fire Pump 4
11.
PS-612 Diesel Fire Pump Start 2-5 Ilours 2x 10 PS-615 Electric Fire Pump Start Ch Ch
--ee LT1
3 Integrated Dose Over Equipment Function Time Needed Time Needed (Rads) 5 12.
SV-4919 Fuel to Diesel Fire Pump 30 Days 1x 10 SV-4918 Fuel to Diesel Fire Pump SV-4935 Cooling Water to Diesel Fire Pump PCV-4515 Cooling Water to Diesel Fire Pump PS-680 I.ow Fire System Pressure PS-789 Through PS-796 Fire Pump Discharge Pressure 5
6 13.
PT-173 Reactor Building Pressure 30 Days 4 x 10 to 1.1 x 10 PT-174 PT-187 5
14.
RDS Channel 30 Days 1x 10 A-D Sensor and Actuator Cabinets 3
15.
125 V D-C MCC 30 Days 1x 10
- 1 and #2 Transformers 5
16.
D01 125 V D-C MCC 30 Days 1 x 10 D02, DIO 125 V D-C Distribution Panel D03 Station Battery Charger Switchgear Station Battery 4
17.
IIPS A-D Uninterruptible Power Supply 30 Days 3x 10 4
18.
P2A Core Spray Pumps 30 Days 3 x 10 P2B 4
4 19.
PS-638 Core Spray Pump Discharge Pressure 30 Days 1.5 x 10 to 3 x 10 HO-7066 Post Incident System Valves HO-7072 Ch 20.
Diesel Generator, 30 Days 1 x 102 Auxiliaries, Control Panel N
6 Ch 21.
Electrical Penetrations 30 Days 1.1 x 10
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BIG ROCK HIGH POINT VENT DESIGN DESCRIPTION 1.0 Introduction - Consumers Power Company committed to install remotely operated high point vents in the reactor coolant system as requested by NRC letter dated September 13, 1979 2.0 Summary - he required venting capability vill be provided by installation of remotely operable vent valves on each of the emergency condenser tube bundles as illustrated in attached sketch #SK-107 Vent valves will be configured such that a single failure vill neither create a reactor coolant system leak nor prevent venting of at least one tube bundle. Venting and operation of one tube bundle vill provide adequate emergency condensing capability.
3.0 Operation - m e emergency condenser vent valves will be normally closed, solenoid operated valves. The four (k) valves will be controlled from the control room with two (2) hand svitches. Panel lights in the control room vill indicate the position of each valve.
4.0 Power supply - Two (2) power supplies vill be used for valve actuation, 125V DC and 120 V AC.
mis vill allow individual vent lines to be con-trolled by separate power supplies as shown in attached sketet #SKE-107 5.0 Single Failures - Series, parallel valves vill provide the ability to vent at least one of the emergency condenser tube bundles in the event of a single failure. Either vent line can be isolated by closing the valves on the corresponding lines to and from the emergency condenser. mis vill still allow venting and operation of one tube bundle.
6.0 Design - h e valves vill be nuclear certified, Class I, solenoid valves with position indication as described below:
Materials of Construction Valves 316 SS Piping 160 Gauge CS Design Pressure 150$ psia Design Temperature 600 F 0 R Design Exposure 2 x 10 Size 3/4 Inch Vent Capability - h e vents vill be used to discharge noncondensable gases (hydrogen) from the emergency condenser tube bundles into the containment area. he gas vill be vented in the proximity of the emergency condenser to allow cooling and mixing with the containment air. Each of'the two (2) vent lines will have the following capacities.
Pydrogen81000 psia,Shg.6 F 3 7 RCS Vol/Er Hydrogen @AT4 Pres,100 F 0 9 RCS Vol/Hr 1661 184
2 70 Analysis 7.1 Failure of a vent line resulting in a LOCA vould be the same as a small steam line break. Previous analyses are adequate to describe this sit-uation, so no new LOCA analyses are needed.
7.2 The results of an analysis of hydrogen generation following an accident were reported in Consumers Power Company letter dated May h, 1979
'Ihis analysis assumed that all hydrogen was released into containment as soon as it was generated. In view of the lov maximum value for hydrogen con-centration demonstrated by this conservative analysis, no additional analyses based on operation of the emergency condenser tube bundle vent valves need be performed.
8.0 Codes and Standards 8.1 Appendix A 10CFR50 General Design Criteria 8.2 10CFR50.h6 Acceptance Criteria for Energency Core Cooling Systems for' Light Water Power Reactors 8.3 10CFR50.h4 Standards for Combustible cas Control System in Light Water Cooler Power Reactors 8.4 Regulatory Guide 1.7 (Rev 1) Control of Combustible Gas Concentrations in Containment Following a Ioss of Coolant Accident 8.5 Standard Review Plan Section 6.2.5 Combustible Gas Control in Containment 8.6 ANSI 331.1 Power Piping Code 8.7 IEEE-279, 1971 - Criteria for Protection Systems for Nuclear Power Generating Stations IEEE-323, 1974 - Qualifying class IE Equipment for Nuclear Power Generating Stations IEEE-344, 1975 - Recomended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations IEEE-379, 1972 - Trial-Use Guide for the Application of the Single Failure Criterion to Nuclear Power Generating Station Protection Systems IEEE-382 -
Trial-Use Guide for the Type-Test of Class I Electric Valve Operators for Nuclear Power Generating Stations IEEE-383 -
Standard for Type-Test of Class IE Electric Cables, Field Splices and Connections for Nuclear Power Gen-erating Stations IEEE-384 -
Standard criteria for Independence of Class IE Equipment and Circuits 9.0 References 91 NRC Letter Dated September 13, 1979 I6 185 9.2 CP Co Letter Dated October 30, 1979 9.3 CP Co Dvg #0Th0Gh0107 Rev M
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APPENDIX E Considerations for Design of Permanent Technical Support Center 1661 188
1 Consumers Power Company is evaluating several alternatives for a permanent Technical Support Center (TSC). Three locations presently under consideration are:
(1) A new structure east of the reactor building with direct access to the turbine building (shown on the conceptual shielding drawings included in Appendix C), (2) the existing Training Center, and (3) the area immediately adjacent to the control room which has been designated the interim TSC.
Locations (1) and (2) would be untenable after an accident which released significant fractions of the core radionuclide inventory and could not be used until additional shielding is installed as discussed in Appendix C.
Location (3) cannot be upgraded before the additional shielding is installed since upgrading work would necessarily displace the interim TSC and no acceptable location exists to relocate it.
Accordingly, a permanent TSC will be incorporated into the design of the additional shield to be added. The interim TSC will continue to be used until the permanent TSC is available.
The permanent TSC will be designed to accommodate 25 persons (nominally),
required plant engineering records and data, and status monitoring and communications instrumentation and equipment.
Instrumentation in the TSC will be capable of providing "real-time" displays of vital plant parameters with recording.
Instrumentation and status monitoring equipment will be nonsafety grade, but comparable in quality to that installed in the control room.
Instrumentation in the TSC will have no effect on the control room; an isolation rack will be installed near the control room - o contain signal isolators, relays, signal converters and test jacks. The isolation rack will be classified Seismic Category I, Safety Class 2.
The status monitoring panel in the TSC will include the necessary information to enable TSC personnel to assess the accident. Consideration will be given to the guidance of Regulatory Guide 1.97, Revision 2, as appropriate to Big Rock Point, but as a minimum the TSC will contain instrumentation to permit assessment of:
a.
Plant safety parameters for:
- Emergency Core Cooling System
- Feedwater System
- Containment b.
Radiological parameters for:
- Containment
- Effluent Treatment
- Release Paths Off-site radiological information including meteorology data.
c.
The TSC instrumentation power supply will not be safety grade, but will be compatible with the requirement to ensure continued availability of informa-tion after the TSC is activated.
~
1661 189
2 In the event a new structure is built to house the TSC, it will not be designed to Seismic Category I standards. Any new structure will meet the requirements of the Uniform Building Code and consider the effects of natural phenomenon anticipated at the site.
The structural design requirements of any new TSC and specifically the need for it to support local shielding is dependent upon resolution of the post-incident shielding issue discussed in Appendix C.
The final design of the permanent TSC will provide personnel protection from radiological hazards due to airborne contaminants in addition to direct radiation. Permanent systems will be provided to continuously monitor radiation levels and airborne radioactivity concentrations in the TSC.
A preliminary estimate of the cost for the TSC, based on a conceptual design, indicates a total investment required of $2 million to $3 million (including required instrumentation).
I66i 190
APPENDIX F Containment Penetrations and Testing Requirements 1661 191
Big Rock Point containment penetrations are tabulated below. This tabulation demonstrates that systems which might contain high levels of.cadioactivity after an accident generally remain inside containment.
Penetrations H-28 H-29, H-112 and H-113 are used for the core spray recirculation system. This system is likely to contain high levels of radioactivity in the event of an accident involving significant core damage.
Periodic leakage testing and preventive maintenance have been implemented on piping of the system outside containment.
1661 192
1 CONTAINMENT PENETRATIONS AND TESTING REQUIREFENTS Service and Testing Required No Penetration Size (Fluid: Frequencv)
Comments H-1 Equipment Lock (12'0")
Air: 6 Month & ILRT Note 1 H-2 Personnel Lock (7'7")
Air: 6 Month & ILRT Note 1 H-3 Escape Lock (3'6")
Air: 6 Month & ILRT Note 1 H4 Manhole (24")
None Manhole on top is welded; FHSR 3.4.
H-5 Eliminated H-6 Eliminated H-7 Ventilation Supply (24")
Air: 6 Month & ILRT Note 1 H-8 Ventilation Exhaust (24")
Air: 6 Month & ILRT Note 1 H-9 Emergency Condenser Air: Refueling & ILRT Emergency condenser sample Vent (24")
point isolation valves and emergency condenser integrity tested during the ILRT.
During power operation, the shell side water is sampled for contamination every 30 days.
H-10 12" Main Steam (24")
Water: NSSS Hydro Test Note 2, see Ref 1 Air: ILRT H-11 10" Reactor Feedwater (20")
Air: Refueling & ILRT Note 2, See Ref 1 H-12 10" Service Water Return None, Not Subject to See Ref 1 (12")
Rupture H-13 10" Service Water (12")
None, Not Subject to See Ref 1 Rupture H-14 4" Space Htg Supply (12")
None, Not Subject to See Ref 1 Rupture H-15 2" Dirty Encl Sump Air:
ILRT See Ref 1, Note 2 1661 193
2 Service and Testing Required No Penetration Size (Fluid: Frequency)
Comments H-16 Spare (4")
None, Tested With ILRT See Ref 1 H-17 Treated Waste Return Air: Refueling & ILRT C:e Ref 1, Note 2 H-18 2" Demineralized Wate-Air: Refueling & ILRT See Ref 1, Note 2 H-19 3" Space Htg Return (6")
None, Not Subject to See Ref 1 Rupture H-20 2" Instrument Air (2")
None, Not Subject to See Ref 1 Rupture H-21 2" Clean Encl Sump Air: Refueling & ILRT See Ref 1, Note 2 Discharge (2")
H-22 2" Reactor and Fuel Air: Refueling & ILRT See Ref 1, Note 2 Pit Drain (2")
H-23 2" Resin Sluice (2")
Air: Refueling & ILRT See Ref 1, Note 2 H-24 Spare (2")
None, Tested During ILRT See Ref 1 H-25 2" Service Air (2"1 None, Not Subject to See Ref 1 Rupture H-26 Spare (6")
None, Tested During ILRT See Ref 1 H-27 Post-Incident Backup (4")
None, Fire Water to Core Is See Ref 1 Contained in Line. Not Local Leak Rate Tested.
H-28 6" Core Spray Pump None, Contains Fire Water See Ref 1 Return (6")
to Core. Tested Each H-29 Refueling Outage and All Observed Leakage Corrected To Zero Leakage Conditions H-30 Spare (12")
None, Tested During ILRT See Ref 1 H-31 Inhibitor Recire Line None, Tested During ILRT See Ref 1 (Shut Down Flush Line)
(12")
H-32 Spares (10")
None, Tested During ILRT See Ref 1 H-33 1661 194
3 Service and Testing Required No Penetration Size (Fluid: Frequency)
Comments H-34 Spare (8")
None, Tested During ILRT See Ref 1 H-35 Control Rod Drive Water: Refueling See Ref 1 Supply (8")
H-36 Post-Incident and None, Contains Fire Water See Ref 1 Fire Supply (6")
To Core. Not Local Leak Rate Tested H-37 Main Steam Drain Water: NSSS Hydro Test See Ref 1 (Nonautomatic Valve)
Air:
ILRT (6")
H-38 Spare (4")
None, Tested During ILRT See Ref 1 H-39 H-40 Electrical Penetration (RDS) Air: Refueling & ILRT See Ref 2, Note 2 H-41 Electrical Penetration None Tested During ILRT See Ref 1 Thru H-64 H-65 See H-40 H-66 See H-41 Thru H-79 H-80 ILRT Pressurizing Line Air: Refueling See Ref 1 H-81 See H-40 H-82 See H-41 H-83 See H-40 H-84 See H-41 Thru H-87 H-88 Gas Meter Connection for Air: Refueling Note 2 ILRT (3/4")
H-89 PS-7064A (3/4")
Note 3 See Ref 1 H-90 PS-7064B (3/4")
Note 3 See Ref I H-91 ILRT, Test Connection Air: Refueling &
See Ref 1, Note 2 (3/4")
Af ter ILRT H-92 Control Air to CV-4029 (3/4") None, Tested During ILRT See Ref 1 66 195
4 Service and Testing Required No Penetration Size (Fluid: Frequency)
Comments H-93 Control Air to CV-4040 (3/4") None, Tested During ILRT See Ref 1, Note 4 H-94 Control Air to CV-4114 (3/4") None, Tested During ILRT See Ref 1, Note 4 H-95 ILRT, Reference Vessel None, Closed to See Ref 1 System (3/4")
Containment Environs H-96 PT-173 (3/4")
Note 3 See Ref 1 H-97 Ventilation Probe Air: Refueling & ILRT See Ref I dpc-9071 (3/4")
H-98 PS-636, -665, -667 Note 3 See Ref 1 PT-174, -187 (3/4")
H-99 PS-637, -664, -666 (3/4")
Note 3 See Ref 1 H-100 Equipment Lock Air:
6 Month & ILRT Note 1 Thru Electrical Penetrations H-103 H-104 Personnel Lock Electrical Air:
6 Month & ILRT Note 1 Thru Penetrations H-107 H-108 Escape Lock Electrical Air: 6 Month & ILRT Note 1 Thru Penetrations H-111 H-112 Core Spray Pump Vent (4")
See H-28 and H-29 See Ref 1 H-113 Core Spray Pump Discharge See H-28 and H-29 See Ref 1 (4")
H-114 Erection Manhole None, Tested During ILRT Manhole is welded shut FHSR 3.4.
Note 1: Semiannual leak rate testing performed with 10 CFR 50, Appendix J, as basis. Also tested during the containment ILRT in the position the valve would be in after an accident.
Note 2: Testing performed during a shutdow. for refueling with 10 CFR 50, Appendix J, as basis. Also tested during the containment ILRT in the position the valve would be in after an accident.
I66i i96
s 5
Note 3:
Instrument connections, one time air test and no leakage observed; closed system.
Note 4:
Control air signal; closed to containment building environs.
References 1.
Consumers Power Company letter dated February 13, 1976.
2.
Consumers Power Company letter dated September 15, 1975.
1661 197