ML19305B329
| ML19305B329 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/14/1980 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8003190534 | |
| Download: ML19305B329 (68) | |
Text
p:::y w} COR80m8f5 a
0 l Power
.- j company y
bgy g
u r
General Offices: 212 West Michigan Avenue, Jackson, Michigan 40201. Area Co.e 517 7HH-OSSO J
March 14, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SHORT-TERM LESSONS LEARNED FROM THREE MILE ISLAND ACCIDENT: ADDITIONAL INFORMATION REQUESTED DURING SITE VISIT An NRC inspection team under the direction of Mr Charles Long visited Big Rock Point February 28 and 29, 1980.
This team reviewed actions taken at Big Rock Point in response to NRC requirements resulting from review of the accident at Three Mile Island (TMI). As a result of this review, Consumers Power Company agreed to provide certain additional information and documentation. to this letter provides responses to the specific items of additional information requested during the NRC site visit. provides Revision 2 to the enclosure of Consumers Power Company's December 27, 1979 letter describing actions to be taken in response to post-TMI requirements.
(Revision I was submitted by Consumers Power Company letter dated January 18, 1980.) Revision 2 updates our response to include the additional information discussed in Enclosure 1 to this letter. Revision 2 also makes minor changes reflecting the completion of certain actions committed in the original response, and incorporates deferral of certain modifications pending a plant risk assessment as discussed in Consumers Power Company letter dated February 22, 1980.
David P Hoffman (Signed)
David P Hoffman Nuclear Licensing Administrator CC JGKeppler, USNRC Enclosures (1) Additional information requested during NRC visit to BRP February 28-29, 1980.
(2) Rev 2 to December 27, 1979 CP Co letter.
80031 DOS 5<f
1 ADDITIONAL INFORMATION REQUESTED DURING NRC VISIT TO BIG ROCK POINT, February 28-29, 1980 NUREG 0578, Item 2.1.2:
" Performance Testing for BWR and PWR Relief and Safety Valves."
Requested Information Big Rock Point design is unique among GE boiling water reactors with respect to relief and safety valves. Clarify, in light of this uniqueness, which of the ongoing valve testing programs Consumers Power Company will participate in.
Response
Based upon Consumers' Power Company's evaluation of the two ongoing valve testing program, that being conducted by the Electric Power Research Institute (EPRI), appears most appropriate. Consumers Power Company is pursuing this matter with EPRI to determine whether Big Rock Point can be enveloped by the EPRI test program.
NUREG 0578, Item 2.1.3.a:
" Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWR's and BWR's."
Requested Information Provide the date by which environmental qualification of the installed equipment will be completed.
Response
Babcock & Wilcox, supplier of this equipment, has initiated a qualification program. This program is a joint effort among B&W and the owners of plants who installed this equipment in response to Item 2.1.3.a.
Consumers Power Company expects this program to be completed by October 1980.
Requested Information Describe the equipment for direct indication of reactor depressurization system (RDS) valve position.
Response
The reactor depressurization system' consists of four valve traias, each containing an isolation valve (gate valve) and a depressurizing valve (pilot-operated relief valve). The position of all eight valves is displayed on the RDS control panel in the control room.
Isolatiod valve position is detected by two limit switches mounted on the valve and actuated by a plate affixed to the piston / valve stem connecting rod. Depressurizing valve position is
2 detected by reed switches actuated by a magnetic indicator rod attached to the valve stem.
NUREG 0578, Item 2.1.3.b:
" Instrumentation for Detection of Inadequate Core Cooling in PWR's and BWR's."
Requested Information Document the capability of switching core spray flow transmitters between flow devices.
Response
NRC letter dated June 4, 1976 required that electrical switching capability be provided. (outside containment) to enable connecting either the ring spray flow transmitter or the nozzle spray flow transmitter to either spray line flow instrument channel. A facility change installing such switching capability was accomplished during 1976.
NUREG 0578, Item 2.1.4:
" Containment Isolation Provisions for PWR's and BWR's."
Requested Information Identify the nine valves for which the modification was made to prevent re-opening when the isolation signal is removed.
Response
The nine valves are:
CV-4031 and CV-4102; enclosure clean sump discharge.
CV-4025 and CV-4103; enclosure dirty sump discharge.
CV-4091, CV-4092, and CV-4093; clean-up demineralizer resin sluice line.
CV-4027 and CV-4117; reactor and fuel pit drain.
Requested Information List the nonessential systems which penetrate containment.
Indicate which signals isolate each penetration.
Response
A tabulation of nonessential systems including the requeste'd information has been incorporated in Rev 2 to our December 27, 1979 letter describing actions taken in response to post-TMI requirements. This tabulation is included in to this letter as Appendix G.
I i
3 NUREG 0578, Item 2.1.5.c:
" Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant."
Requested Information Document the maximum expected containment hydrogen concentration using the assumptions specified in Regulatory Guide 1.7.
Response
Regulatory Guide 1.7, and 10 CFR 50.44, specify that analysis of post-accident containment hydrogen concentration should assume generation of five times the amount of hydrogen predicted by analyses demonstrating compliance with 10 CFR 50.46 (but in no case less than that which would result from reaction of all the metal in the outside surfaces of the cladding cylinders surrounding the fuel to a depth of 0.00023 inches). Big Rock Point analyses demonstrating compliance with 10 CFR 50.46 predict less than 1% metal-water reaction. The amount of cladding corresponding to a depth of 0.00023 inches is 0.68% of the total cladding. Regulatory Guide 1.7 assumptions, therefore, dictate consid-eration of the hydrogen which would result from a metal-water reaction involving 5% of the total cladding.
A metal-water reaction involving 5% of the total cladding, followed by uniform dispersal of the evolved hydrogen in the containment atmosphere, would result in a containment hydrogen concentration of 0.3 v/o. This analysis assumes' essential'.y no hydrogen is dissolved in the coolant at the start of the acci-dent which is a reasonable best estimate assumption for a BWR.
Increases in containment hydrogen concentration from poct-accident radiolysis are conserva-tively predicted to add 1 v/o in approximately two months. Over one year would be required for radiolysis to increase the concentration to the value considered flammable (4 v/o).
NUREG 0578, Item 2.1.6.a:
" Integrity of Systems Outside Containment Likely To Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWR's and BWR's."
Requested Information Describe the procedure used for leak testing post-incident core spray recircu-lation lines.
Response
Leak testing is accomplished during the core spray pump operability test.
This test operates the core spray pumps to recirculate water from a test tank inside. containment. The test" consists of changing valve lineups so the core spray pumps draw from the test tank rather than the containment strainers and discharge to the test tank rather than to the core sprays. Each pump is operated at full discharge pressure. Piping, pumps, and valves are visually observed for leakage..A maintenance order is issued if any leakage is.
o 4
observed to return the system to a leaktight condition. This test is performed during each refueling outage.
Requested Information Report the results of review of the plant in light of information provided in NRC letter dated October 17, 1979 and IE Circular 79-21 describing an incident at North Anna.
Response
The North Anna incident was related to a construction error and Item 2 of the circular is related to design and construction concerns. A total of 7 incidents of inadvertent / unplanned releases have occurred in our 17-year operating history. These have been rereviewed. None of these incidents involved errors in construction. Four involved component defects which have been addressed to prevent recurrence. A detailed review of release pathways has not been done specifically for this item but our operation and review activities associated with 17 years of operation provide adequate assurance that the plant configuration as related to radiological control is as shown on the controlled drawings. A 1975 floor drain verification has been updated.
Item 1 of the circular is related to review of procedures for transfer of radioactive liquids. Our operating history indicates that four minor unplanned off-site releases involving liquids and gas have occurred that are partially or entirely related to administrative control problems. Our action to prevent recurrence has proven adequate for these incidents and existing procedures and controls are deemed adequate tc minimize inadvertent releases.
Item 3 of the circular is related to testing and verification of systems that could cause an inadvertent release. Systems do exist that could become
-pathways for inadvertent release. These include:
1.
Condensate tank vent to outside atmosphere.
2.
Condensate tank and piping in soil bed.
i 3.
Radwaste discharge pipe in soil bed.
4.
Drain line from post-incident room to radwaste in soil bed.
' Of all these items, only one (Item 4) is associated with the system (post-incident system outside containment) identified in our December 27, 1979 letter as requiring testing for Item 2.1.6.a.
The elimination of leakage through this testing will eliminate the presence of radioactive liquids in this line.
5 NUREG 0578, Item 2.1.6.b:
" Post-Incident Shielding."
Requested Ir. formation Provide radiation maps of the plant site (ie, inside the security fence) after an accident involving a release of fission products in the amounts specified in NUREG 0578. Describe the basis for the dose calculations.
Response
Radiation maps have been added to Appendix C of the enclosure to our December 27, 1979 letter in Revision 2.
Revision 2 is forwarded as Enclosure 2 to this letter. The calculations on which these maps were based were performed as follows:
10 3 1.
Containment free volume above ground level (2.48 x 10 cm ) was divided into 71 equally sized volume segments. The center of each segment was located by an orthogonal coordinate system having its center on the vertical axis of containment at ground level (elevation 592.5 feet above sea level). Location of each volume segment within containment free volume (excluding major internal structures) was aided by use of a scale model of the plant and balloons representing the volume segments.
2.
Twenty positions outside containment of special interest for dose rate determination were also located. Distances from each " visible" volume segment to each of the 20 positions were determined. A volume segment was' considered " visible" if less than 2 feet of concrete shielding was present between the volume segment and the dose receptor point. A virtual source distance then was calculated (square root of the sum of the distances divided by the number of visible volume segments). This represents the distance at which one volume segment, if it contained the activity of all visible segments, would give the same dose rate as all visible segments at their actual distances.
)
3.
Use of the virtual distance calculation method was found to be necessary only for dose receptor points within the containment or at the contain-ment surface. For all other locations, acceptable accuracies (within 10%
of virtual distance results) were obtained by using an 1852 cm radius sphere which represents the containment free volume. This sphere was assumed to contain a gaseous source term at a concentration adjusted by the ratio of " visible" to total volume segments. A computer program was utilized for these calculations which sums dose rates at any designated position within or outside of a spherical source. Dose rates for each gamma emission of each airborne nuclide are summed. Various shields, including the 0.75-inch steel containment shell and intervening thick-nesses of concrete or other materials, are considered by the program in determining dose rates at the desired loca-tions.
6 4.
Approximate isodose lines were drawn based on calculations for several time intervals af ter the assumed release of a TID 14844 source term into the containment atmosphere. Source terms were calculated by ORIGEN (run SNUMB = 70075, dated November 9, 1979) as provided by the BWR Owner's Group.
Requested Information List the items in the plant that will be evaluated for equipment protection during 1980.
Response
The items to be evaluated are indicated by an asterisk in-the Revision 2
. tabulation of. equipment to be evaluated. This tabulation is included in to this letter. Electrical equipment so indicated is involved in a current SEP evaluation of equipment qualification which is expected to be complete by May 15, 1980. The effort which must be applied to this SEP evaluation precludes concurrent evaluation of mechanical equipment.
Mechanical equipment evaluation will be complete by July 15, 1980. Equipment found deficient in these evaluations will be modified / replaced by January 1, 1981 to the extent possible.
NUREG 0578, Item 2.1.8.b:
" Increased Range of Radiation Monitors."
Requested Information Modify the range of the long-term effluent gas monitor to 10' uCi/cc.
Response
Revision 2 attached incorporates this change.
Requested Information Clarify the criteria to which each of the long-term instruments will be designed.
Response
This clarification is incorporated in Revision 2.
Requested Information
' Modify the commitment for installation of long-term iodine sampling of plant effluents to reflect completion by January 1, 1981.
Provide the date by which a detailed design of the proposed system will be submitted for NRC review.
4
-r_,_
7
Response
This commitment is modified as requested in Revision 2 attached. A design description will be submitted by July 1, 1980.
~NUREG 0578, Item 2.1.8.c:
" Improved In-Plant Iodine Instrumentation."
Requested Information Document plans to upgrade capabilities in the control room / technical support center area beyond that previously proposed.
Provide a date for completion.
Response
-Portable spectral analyzers will be purchased to improve this iodine monitoring capability. These instruments are expected to be available by November 1, 1980.
Requeste'd Information 1
Revise the docketed response to reflect c<.opletion of actions now done.
Response
This is incorporated in Revision 2.
1 NUREG 0578, Item 2.2.1.b:
Requested Information Provide a description of the selection and training program for shif t techni-cal advisors.
Identify training already accomplished and any planned.
1
Response
Candidates for shift technical advisor (STA) are required to have a bachelor's degree or equivalent in a scientific or engineering discipline (all current STAS are degreed engineers). Candidates for replacement STAS will be expected to commit to two years in this position. The training program for STAS in-cludes the subjects listed below; nuclear plant experience may serve to reduce
-or_ replace certain aspects of this training provided the basis for such waiver is documented.
The training program for STAS includes:
1.
Plant General Employee Training 2.
Duties and Responsibilities of an STA 3.
Quality Assurance Indoctrination-
8 4.
Plant Engineering Manual 5.
Radiation Work Permit Exemption and Respirator Qualification 6.
Basics of Nuclear Physics and Reactor Operating Response, Thermodynamics 4
7.
Plant Design and Layout (FHSR) 8.
Plant System Descriptions and Standard Operating Procedures 9.
Plant License and Technical Specifications 10.
Plant Administrative Procedures and Security Plan 11.
Analyzed Accidents (FHSR) 12.
Transient Responses and Common Mode Failure Analysis 13.
Plant Emergency and Off-Normal Procedures 4
14.
Health Physics Procedures: Sampling, Releases, Release Evaluations, Radwaste Transport 15.
Site Emergency Plan and Implementing Procedures 16.
Accident Assessment and Response 17.
Operating Experience Engineering Evaluation 18.
Simulator Training: Casualty Responses 19.
Management and Supervisory Skills Requalification training for STAS will include the following material in addition to that required as part of General Employee Training:
1.
Review of Facility Changes / Specification Field Changes i
2.
Review of Applicable Procedure Changes
- 3. ' Review of Plant and Industry Licensee Event Reports 4.
Review of Heelth Physics Procedures for RWP Exemption 5.
Review of Revisions to Technical Specifications Tests 6.
Review of Changes to Engineering Manual 7.
Review of Revised Accident and/or Transient Analyses
_. _.,, ~
cip
9 8.
Simulator: Casualty Responses 9.
Review of Technical Specifications Changes Training already completed for current STAS includes:
1.
Orientation 2.
RWP Exemption 3.
Security 4.
Site Emergency Plan 5.
Technical Specifications 6.
Operating Procedures 7.
Final Hazards Summary Report 8.
Thermodynamics 9.
Administrative Procedures 10.
DOT-NRC Waste Handling Requirements 11.
Health Physics Procedures - Inadvertent High Radiation E,osure 12.
Simulator:
Casualty Responser NUREG 0578, Item 2.2.2.b:
" Control Room Access."
Requested Information Revise procedures and previous responses to remove the STA from operational command during an accident. Correct the Management directive which allows this situation.
Response
Consumers Power Company procedures, previous responses, and Management directives require no change; existing documents clearly define the command authority as residing with the shift supervisor or the No 1 control operator.
The STA never is in a position of command responsibility and has no command authority. The procedures do permit the shift supervisor to leave the control room when necessary during emergency situations and specify that command authority passes to the No 1 control operator; the STA is required to remain in the control room under these circumstances so that he will be available to provide technical advice to the No 1 control operator, if requested.
10 The requirement for the shif t supervisor to remain in the control room during accident situations has been the subject of correspondence between the BWR Owners Group'and the NRC. Specifically, Owners Group letter dated October 17, 1979 (in response to requirement 2.2.1.a) stated that the shif t supervisor would remain in the control room whenever a site or general emergency has been declared. NRC letter dated November 14, 1979 accepted this interpretation.
Consumers Power Company will revise procedures to ensure that they are consistent with this BWR Owners Group position.
ACRS Requirement: " Hydrogen Monitoring."
Requested Information Clarify Consumer Power Company's position on installing hydrogen monitoring in light of Consumers Yower Company's letter dated February 22, 1980.
Response
Consumers Power Company's letter dated February 22, 1980 discussed performance of an integrated plant risk assessmen*;. Provision of a containment hydrogen concentration monitor was among the list of modifications to be deferred pending completion of the risk assessment. This has been clarified in Revision 2.
g IMPLEMENTATION CRITERIA FOR POST-TMI REQUIRUENTS
~ g BIG ROCK POINT LIST OF EFFECTIVE PAGES Page Number Revision 1
Original 2
1 3-12 2
13-lh original 15-17 2
18-21 Original 22-23 1
2h 2
25 Original 26-27 2
27a 2
28 2
29-31 1
32 2
33 1
3h-37 2
38 Original 39 1
h0 h1 2
kla, b & c 2
h2 h3 2
hk Original 45 h6 2
h7 Original h8 1
49 2
50-5h Original 55-56 2
5T Original 58 1
Rev 2 3/1h/30
e i
4 4
2 List of Effective Pages - continued d
Page Number-Revision i
59-62 Original Appendices Original (except as i
follows)
I. ~
Appendix C (except 2
Alternative Shield Design Drawings)
Appendix G 2
i 4
s Rev 2 3/lk/80 r
e p
e
., - + +,
,.~,v
.r.-,--n.+-.., =-, -
---w...-sw--,nw- - - -
3
'2.
NUREG 0578 Requirement 2.1.2:
" Performance Testing for BWR and PWR Relief and Safety Valves."
Commit to provide performance verification by full scale prototypical testing for all relief and safety valves. Test conditions shall include two-phase slug flow and subcooled liquid flow calculated to occur for design-basis transients and accidents.
l Action To Be Taken at Big Rock Point Analyses of Big Rock Point transient response have consistently indicated 1
l that the safety valves would not be required to pass two phase flow for any design basis transient or accident. In fact, operation of even a single loop j
of the emergency condenser serves to preclude opening of a safety valve for essentially all analyzed events. ' Safety valve cycling is, therefore, not.a routine occurrence at Big Rock Point.
Big Rock Point's design is unique among GE BWRs with respect to safety / relief valves. Only spring safety valves are provided.
In addition, Big Rock i
. Point's higher operating pressure results in safety valve set points considerably higher than for other BWRs. For these reasons, it appears that l
the BWR Owners Group safety / relief valve test program is inappropriate for Big Rock Point. Consumers Power Company is, therefore, pursuing negotations with the Electric Power Research Institute to determine whether Big Rock Point can l
be enveloped within the EPRI test program.
l Previous correspondence has identified that the BWR Owners Group was considering installation of high level feedwater trip systems as a means of Rev 2, 3/14/80 j
4 i
addressing the concerns of'this requirement. The Owners Group has recently concluded that'such a recommendation is inappropriate. Consumers Power Company has terminated its efforts toward design and installation of such a trip system.
t i
4 4
4 l
4 1
1
.I
'f l
l 3
-(
Rev 2, 3/14/80 i
-w--
y r-m v-+
w
,--w
~
e
5 3.
NUREG 0578 Requirement 2.1.3.a:
" Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWR's and BWR's."
Provide in the control room either a reliable, direct position indication for the valves or a reliable flow indication device downstream of the valves, i
l Action To Be Taken at Big Rock Point i
l 1
Indication will be provided for each of the six safety valves. Acoustic sen-sors will be used for this purpose. A single sensor will be associated with each valve. The installation will meet the,same requirements as other engi-neered safety features (except redundancy). Equipment which can be made t
available to support rapid installation has not all been fully certified to post-accident environmental conditions. Babcock & Wilcox, equipment supplier, has initiated a certification program in cooperation with owners who utilized their equipment for this purpose. This certification program is expected to be completed by October 1980.
Appendix A to this. enclosure includes, for NRC information, a detailed de-scription of the safety valve position indicator system to be installed. This appendix includes the following documents prepared by the design contractor, Energy Incorporated: project requirements, construction specifications for the portion of the system inside containment, construction specifications for the portion outside containment, and the test specification for the system.
7 Big Rock Point was removed from service prior to January 1, 1980 to install the position indicator system described in Appendix A.
The system installa-
> tion was completed and the system was energized prior to plant start-up. The Rev 2, 3/14/80 d
I -
o
6 system is now operable, following final adjustment for correct bandwidth after full power conditions had been achieved.
Backup methods for determining that a safety relief valve is open include a common drain header high temperature alarm, contaiment high pressure alarm, and indication and dewcell temperature recorder for high humidity and pipeway high temperature. These indirect methods of recognizing an open valve are discussed in Plant Operating Procedures.
Reactor depressurization system (RDS) valves are presently equipped with direct position indication in the control room.
Isolation valve position is detected by two limit switches actuated by a plate affixed to the piston / valve stem connecting rod. Depressurizing valve position is detected by reed switches actuated by a magnetic indicator rod attached to the valve stem. The position (open or closed) of each valve is indicated on the RDS control panel in the control room.
i -
J Rev 2, 3/14/80 i
7 4.
NUREG 0578 Requirement 2.1.3.b:
" Instrumentation for Detection of Inadequate Core Cooling in PWR's and BWR's."
Perform analyses and implement procedures and training for prompt recognition of low reactor coolant level and inadequate core cooling using existing reactor instrumentation (flow, temperature, power, etc) or short-term modifications of existing' instruments. Describe further measures and provide supporting analyses that will yield more direct indication of low reactor coolant level and inadequate core cooling such as reactor vessel water level instrumentation.
Action To Be Taken at Big Rock Point Consumers Power Company is participating in the BWR Owners Group effort in this area. Analyses performed by General Electric Company for the Owners Group have indicated that adequate core cooling can be assured by verifying either a vessel water level above the core or rated flow in one of the low-pressure core spray systems. Big Rock Point is equipped with a reactor vessel level instrument and flow indication in each of the two core spray lines.
j Electrical switching capability exists outside containment to enable either spray line flow transmitter to be connected to each spray line flow instrument-channel'. Procedural guidelines addressing the use of these instruments are under development by the Owners Group and will be implemented on the schedule agreed between the Owners Group and the NRC.
NRC letter. dated July 17, 1979 (Enclosure 1, Attachment 3) requested information concerning the effects of post-accident conditions on reactor vessel level indication. This issue was addressed in General Electric report Rev 2, 3/14/80 s
-~
- - ~
~
r -
8 NEDO-24708 submitted on behalf of the BWR Owners Group on August 17, 1979.
This area was also discussed in a General Electric Company Service Information Letter. Specific evaluation of this issue for Big Rock Point identified a concern with regard to the ability of the temperature-compensated level instruments to reliably perform their safety functions following a postulated accident; this concern was reported by Licensee Event Report (LER) 79-22 l
2 submitted September 5, 1979. The existing level instrumentation was modified to eliminate temperature compensation and alleviate the concerns discussed in l
LER 79-22.
Consumers Power Company lettars dated October 23 and October 31, 1979 discussed the modifications in support of Technical Specifications changes which they necessitated; the revised Technical Specifications were issued as Amendment No 31 to Licedse DPR-6 on November 2, 1979.
Due to the differences between reactor vessel / steam drum level instrumentation i
available at Big Rock Point and that in newer BWRs, Consumers Power Company is evaluating alternate level instrumentation designs. Energy Incorporated (EI) i is currently engaged in designing new level instruments. A primary objective of new instrumentation is to provide accurate indication and trip signals during normal operation, anticipated transient and accident conditions, and i
l post-ac ident conditions. To achieve this objective, the new system design will be selected to further reduce the effects of post-accident containment heating on the reliability of indication. The new system will have a range l
equal to that of the present instrumentation.
.In addition to replacing the existing instrumentation with a design'of improved reliability, consideration was given to installing a wide-range level indication. Big Rock Point's steam drum and riser design proved to be a Rev 2, 3/14/80
9 significant obstacle to providing the desired wide-range coverage. Further consideration of additional level instrumentation is being deferred pending completion of an overall risk assessment as discussed in Consumers Power Company-letter dated February 22, 1980.
EI has completed conceptual design of a new level instrumentation system incorporating the features discussed above. A detailed design description of the instrumentation selected will be submitted for NRC information by July 1, 1980. The new instrumentation will be installed by January 1, 1981.
1 J
Rev 2, 3/14/80
. _. ~.,
_ - ~ _ _ _ _ _
10 5.
NUREG 0578 Requirement 2.1.4:
" Containment Isolation Provisions for PWR's and BWR's."
Provide containment isolation on diverse signals in conformance with Section 6.2.4 of the Standard Review Plan,, review isolation provisions for non-essential systems and revise as necessary, and modify containment isolation designs as necessary to eliminate the potential for inadvertent reopening upon reset of the isolation signal.
Action To Be Taken at Big Rock Point Diversity currently exists in parameters sensed for containment isolation.
Either low reactor vessel water level or high containment pressure results in an isolation. Normally open nonessential lines which carry fluids outside containment are closed automatically on an i. solation signal. Normally open lines carrying fluid into containment are equipped with check valves and can also be secured by manually operated gate valves or air-operated control valves. Systems considered essential are as follows:
Post-Incident and Fire Water Supply System j
Post-Incident Backup System Ventilating Vacuum Breaker Sensing Line j
Core Spray Recirculation System i
{
Further information regarding nonessential systems which penetrate containment ir, provided in Appendix G.
Classification of systems,as essential /non-
- essential will be reviewed as part of SEP.
Rev 2, 3/14/80 t
n
~
r
11 Consumers Power Company letter dated May 4, 1979 provided our response to IE Bulletin 79-08.
This response stated, in part, that certain valves closed by a containment isolation signal could automatically reopen if the sctuating signal cleared; interim administrative actions to prevent this were described, and a commitment was made to modify the isolation logic to preclude such automatic reopening. The problem of inadvertent reopening of the isolation valves was found in four (4) control circui* which control nine (9) isolation valves. The nine valves are CV-4031, CV-4102 (enclosure clean sump discharge), CV-4025, CV-4103 (enclosure dirty sump discharge), CV-4091, CV-4092, CV-4093 (clean-up demineralizer resin sluice)', CV-4027 and CV-4117 (reactor and fuel pit drain). The automatic reopening of the isolation valves was eliminated through the addition of a seal-in relay and a push-button switch into each of the four (4) control circuits.
Both the seal-in relays (General Electric Type HMA relay) and the push-button switches (General Electric industrial miniature oil-tight push button) are of a quality equal to or greater than similar equipment presently in use in Big Rock Point safety systems. The seal-in relay was installed in each isolation valve control
}
circuit before the isolation valve hand switch and in parallel with the circuit. One normally open contact of the seal-in relay was placed in series l
with the circuit and before the seal-in relay coil to act as a holding contact. This arrangement is shown diagrammatically in Appendix B.
A push-button switch, with momentary contacts, was installed so as to bypass the holding contacts. This provides a means of energizing the seal-in relay coil and closing the holding contact which, in turn, completes the circuit, provided that the reactor protection systems containment isolation signal is clear and the isolation valve's hand switch is in the closed position.
If Rev 2, 3/14/80
12 loss of a-c power should occur or reactor protection systems containment
-isolation signal should actuate, the seal-in relays coil will become de-energized, dropping out the contact.
(See A.pendix B.)
Re-energizing of the circuit can only be accomplished through a direct operator action via the push-button switch, and only after the de-energizing signal is corrected or cleared.
In conjunction with the seal-in relay and the push-button switch, an indicating lamp was added in parallel with each isolation valve control circuit to indicate when the holding contact is made up.
The system described above was installed during the outage discussed in response to Requirement 2.1.3.a above.
Installation was completed before the first reactor start-up of 1980.
I L
j Rev 2, 3/14/80 t~
15 8.
NUREG 0578 Requirement 2.1.5.c:
" Capability To Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant."
A minority of the Task Force recommended.that all light water reactor plants have the capability to install recombiners within a few days following an accident.
Procedures to accomplish such installation were to be developed.
Action To Be Taken at Big Rock Point Big Rock Point's licensing basis does not include use of hydrogen recombiners. No action will be taken.
Consume ~rs Power Company letter dated May 4, 1979, in response to IE Bulletin 79-08, reported the results of an analysis of the total amount of hydrogen which could be
. generated by a metal-water reaction involving all the cladding in the fueled region of the core. The maximum possible hydrogen concentration was incorrectly reported as 5.7 v/o; the correct value is 6.7 v/o. Despite this error, Consumers Power Company continues to consider that the hydrogen concentration which might be present inside containment is sufficiently small to preclude the need for hydrogen recombiners.
Conservative assumptions used in this analysis included the assumption 1
l that the reactor coolant system contained the maximum possible dissolved hydrogen (ie, was saturated) at the start of the accident.
More realistic assumptions predict a much smaller value for containment hydrogen concentration. Use of the assumptions specified in Regulatory Guide 1.7 (five times the metal-water reaction predicted by 10 CFR 50.46 analyses) and assuming no hydrogen initially dissolved in t.he coolant, a maximum concentration of 0.3 v/o is predicted. Over one year of post-accident radiolysis would be required to raise this concentration to the value considered flammable (4 v/o).
Rev 2, 3/14/80
16 9.
NUREG 0578 Requirement 2.1.6.a:
" Integrity of Systems Outside-
. Containment Likely To Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs."
. A program shall be implemented to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious accident. The program shall include (1) implementing all practical leak reduction metheds, (2) measurement and report to NRC of actual leak rates and (3) preventive maintenance including at least integrated leak tests each t
refueling outage.
Action To Be Taken at Big Rock Point The core spray system would contain radioactivity af ter an accident. The piping and valves in this system which are outside containment will be observed for leakage when testing the core spray pumps.
In addition, core spray heat exchanger tube leakage will be calculated to ensure no tube
\\
degradation in excess of Technical Specifications limits. The core spray pump operability test procedure has been modified to require monitoring for leakage and issuance of a maintenance order for repair if any leakage is observed.
This test consists of operating each core spray pump by recirculating from a test tank inside containment. This test is performed during each refueling outage. This test was performed October 18, 1979 with no observable leakage.
-The radwaste system could contain radioactivity af ter an accident. However it
- would be unlikely that high levels of radioactivity would be found since double valve isolation is provided for all sources of such contamination. The primary concern for off-site exposures from radioactivity in this system would Rev 2, 3/14/80
17 be release of radioactive gases from degasification; leakage testing would not c'ontribute to reduction of off-site doses from such degasification since all tanks :ba the system are open ' to the atmosphere. Thus, the radwaste system will not be_ tested for leakage in response to this requirement.
Big Rock Point design is such that other systems which might contain radioactivity.after an accident are generally inside containment. Systems 5
penetrating containment and the tests performed on each penetration are tabulated in Appendix F.
i Plant design, operating experience, and operating procedures have been-reviewed in view of information contained in NRC letter dated October 17, 1979 and IE Circular 79-21 regarding an incident at North Anna. The conclusion of this review was that no unknown radioactive release paths were identified. The few instances.of inadvertent radioactive releases which have occurred during our seventeen years of operating experience have been reviewed; these instances resulted either from component failure for which corrective action was taken to prevent recurrence or from administrative deficiencies which were 4
promptly corrected.
i r
t-Rev 2, 3/14/80
24 The monitor was in place and procedures for its use as discussed above were completed before start-up from the outage discussed in response to Requirement 2.1.3.a above.
It should be noted that this method would also provide backup information for accidents wherein lesser core damage occurs and the sampling procedure of No I above can be used.
3.
Long-Term Solution Conshmers Power Company's preliminary evaluation of possible methods of providing the required capability in the long term centered around techniques for in-line sample monitoring. This would eliminate the problems attendant upon handling of samples containing high levels of radioactivity which would be involved in laboratory analysis.
Dissolved boron is not used for reactivity control at Big Rock Point except in the liquid poison system; thus, sampling capability to analyze boron is not needed.
Lake Michigan coolant used for heat rejection in the main condenser and as the emergency core cooling water supply is extremely low in chlorides; thus, chloride analysis capability would not enhance post-accident diagnosis. The small size of Big Rock Point's core with i
respect to the large containment free volume results in essentially no l
concern due to hydrogen generation.
(See 2.1.5.c above.) Thus, hydrogen analysis capability is not needed. Quantification of radionuclide content can be performed using an in-line monitor.
It is, therefore, concluded
.that capability to obtain coolant and containment atmosphere samples for laboratory analysis is not needed.
Rev 2, 3/14/80
26 14.
NUREG 0578 Requirement 2.1.8.b:
" Increased Range of Radiation Monitors."
Provide high-range radiation monitors for noble gases in plant ef fluent lines and ' redundant high-range radiation monitors in the containment. Provide capability of measuring and identifying radioiodine and particulate radioactive effluents under accident conditions.
Action To Be Taken at Big Rock Point The following will be provided by January 1, 1981:
Radioactive noble gas effluent monitors from ALARA rabges to 10 Ci/cc.
Capability of radioiodine sampling of effluents followed by on-site analysis.
Conta1nment radiation level monitors (minimum of 2) capable of measuring radiation levels to a maximum of 10 rad /h.
The maximum ranges of the instrumentation to be provided were determined based on an analysis of the highest possible radionuclide release at Big Rock Point.
Big Rock Point's containment is large (approximately 2.66 x 1010 cm3) while core size is considerably smaller than in newer plants. These factors result in maximum post-accident radiation levels and noble gas concentrations (assuming Regulatory Guide 1.3 release fractions) which can be monitored by instruments of the specified ranges.
Instrumentation installed will meet the following criteria:
Rev 2, 3/14'/80 1
1
27 High-Range Effluent Monitors 1.
Effluent monitors are for noble gases only.
2.
Noble gas effluent monitors will not be redundant (one per normal release point).
3.
No seismic qualifications are required.
4.
Power supplied by emergency power bus.
.5.
Continuous display and recording via chart or digital printer.
- 6. ' Regulatory Guide 1.97 and ANSI N320-1978 will be followed when consistent with the specifications above.
High-Range Containment Radiation Monitors 1.
Sensitive to Photon radiation only.
2.
Two physically separate and redundant containment radiation level monitors will-be provided.
5 3.
Seismic. qualifications will be equivalent to the systems to which it is associated.
4.
Power supplied by ' emergency power bus.
5.
Continuous display and recording via chart or digital.rinter.
Rev 2, 3/14/80 n
e
..,w
27a 6.
Laboratory calibration to be supplied by vendor. Electronic calibration will be used for upper end of monitor to preclude high personnel radiation exposures. Large source strength would be necessary for direct calibration.
7.
' Regulatory Guide 1.97 and ANSI N320-1978 will be followed when consistent with the specifications above.
Interim methods of obtaining the required information and additional information regarding the long-range instrumentation to be provided is as follows:
1.
Interim Methods of Quantifying Gaseous Effluents Current stack monitor capabilities are limited to a release rate of approx-imately 40 Ci/s which represents three times the Regulatory Guide 1.3 t
Rev 2, 3/14/80
28 maximum hypothetical accident (MHA) source term with leakage at a rate of 0.5% of gaseous inventory per day and 30,000 cfm stack flow. The current monitor samples a stream which would contain all leakage from turbine building sources, including condenser off gas.
a.
System and Method (1) A monitor with readout which is remote from the detection chamber is being dedicated to emergency use in quantifying high level releases. Dynamic range requirements of approximacel" 5 mR/h to 100 R/h have been determined for chosen sensor location in order to quantify release rates from the level of current instruments and procedures t'o the level of 10,000 Ci/s. Sensitivity of the sensor to Xe-133 (81 kev) will be accounted for in the graph for dose rate-to-release rate conversion. Tables will provide conversion values as a function of time after shutdown such that the larger percentage contribution from Xe-133 at later times will be acknowledged. The installed instrument, tables, and graphical displays of tabulated data, were available for use with plant procedures prior to start-up from the outage discussed in 2.1.3.a above.
(2) Monitor location for interim high level stack release monitoring is adjacent to current stack gas sample and return lines. A lead shield for background reduction is provided. The monitor location has been chosen at a level of approximately 13 feet below grade in order to eliminate direct shine from the unshielded containment structure. Measurements Rev 2, 3/14/80
.y..
w
32 Class B.
The previously calculated X/Q can then be used to calculate an estimated release rate.
The procedures and equipment described above were available for use before start-up.
3.
Noble Gas, Particulate and Iodine Monitoring - Long-Term Methods Iodines and entrained noble gases would provide a source of approximately 800 curies in the charcoal adsorber for a 15-minute (3 cfm) sample following an accidcat with Regulatory Guide 1.3 releases at Big Rock Point. Only one commercially available sampling system (Science Applications /RadeCo) provides the automated (hands off) sampling and analysis which is required for personnel protection at these levels.
Automatic fresh air purge of the charcoal releases the noble ' gas so that a maximum of not more than approximately 300 curies of iodine would be present for automatic analysis by an intrinsic germanium spectrometer.
e
- We are evaluating an SAI proposal for application at Big Rock Point. The SAI system, or its equivalent, will be installed by January 1, 1981 provided equipment can be made available. A design description of the
. proposed system will be submitted by July 1, 1980.
1 i
1 5
Rev 2, 3/14/80
34
- 15. NUREG 0578 Requirement 2.1.8.c:
" Improved In-Plant Iodine Instrumentation."
Each licensee shall provide equipment and associated training and procedures for. accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.
Action To Be Taken at Big Rock Point Airborne radioiodine monitoring with conventional charcoal canisters may produce overly conservative results under accident conditions due to interference from radioactive noble gases. Occupancy of various locations could be unnecessarily restricted if airborne radioactive iodine levels were incorrectly determined to be high.
1.
Short-Term Solution RadeCo Model GY-130 silver zeolite radioiodine sampling cartridges demonstrate an acceptable collection efficiency for inorganic as well as organic iodines (94-96% in the normal sampling flow rate range) but do not retain noble gases to any appreciable degree. Although these cartridges are too expensive ($45 each) for normal plant sampling, they are justifiable for emergency monitoring. Two hundred of these filters have been procurcd.
Respiratory protection is currently available and potassium iodide as a thyroid blocking agent is available to reduce thyroid burdens of radioactive iodine.
Rev 2, 3/14/80
35 e
' Procedural Method a.
A 10-minute (10-20 cubic foot) air sample will be passed through a combination particulate filter and silver zeolite cartridge holder at a flow' rate of I-2 cfm using the standard Big Rock Point air sample, the RadeCo Mocel H809V. The silver zeolite cartridge will then be counted using a otandard frisker (Eberline RM-14, Ludlum 177 or equivalent) equipped with a pancake GM probe. Collection efficiency of 94% and detector efficiency of 25% yield a minimum detectable activity of 6.0E-10 Ci/ml for the smallest sample size (10 cubic feet), which is a factor of 10 below the I-131 MPC. Projected iodine levels in occupied areas in excess of 520 MPC hours (equivalent to 40 MPC hours for 13 weeks permitted by 10 CFR 20.!03) will require respiratory protection to be worn or potassium iodide to be prescribed.
Evaluation of the type of protection to be used will be made on a case-by-case basis depending upon the individual action required in the airborne area. The Company's general medical consultant has developed a procedure for the use of potassium iodide in emergency situations.
Initial draf t of this procedure was submitted to the Company on December-13,.1979 and was implemented by January 15, 1980.
b.
Equipment Impact (1) Without reuse, a quantity of approximately 200 cartridges is adequate to permit sampling of the Control Room / Technical Support Center (TSC) and the Operations Support Center atmospheres as I
Rev 2, 3/14/80
A 36 i.
conditions dictate for the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during an accident and then every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for ' the next 10 days. With reuse of 4
cartridges which have insignificant accumulations of radioactivity, sampling can be expected well in excess of this minimum frequency.
(2) Portable air samplers powered by 120 V a-c emergency bus and survey instruments are available for use.
(3) Eight standard size (2" x 4" x 8" or 2" x 4" x 6") lead
-bricks are stored in both the TSC and in the Operations Support l
i Center such that they are available for the construction of counting shields for pancake GM probes in the event that background-radiation levels preclude air sample cartridge
)
counting without' shielding.
.l c.
Training The sample procedure is similar to existing surveillance air sample.
4 i
procedures and utilizes existing air sampling equipment and counting -
- instruments. Training for health physics personnel has consisted of a required review of the procedure.
A group discussion _on each shift, led by the responsible assistant supervisor, also has been performed.
l Rev 2, 3/14/80
,+t
37 2.
Longer Term Actions I
-Orders have been placed.for_several additional a-c/ battery-operated friskers (Ludlum Model 177-or equivalent ratemeters) to supplement the existing supply and ensure that adequate numbers of friskers will be available for contamination monitoring and air sample counting during an emergency, c
Battery-operated ratemeters (Eberline PRM-6 or equivalent) are being
.l purchased to permit monitoring and air sample counting in the field during emergencies.
Portable spectral analyzers will be procured to improve iodine monitoring capability. These will be available by November 1, 1980.
J 1
1
-l 1
f Rev 2, 3/14/80-
i 40 Shift. supervisor's duties have been reviewed. Duties which detract from or 4
are subordinate to the primary responsibility for safe operation of the plant j
were assumed by another individual. This was accomplished prior to start-up from the outage discussed in 2.1.3.a above. Shift supervisor duties will be
, reviewed on an annual basic by the Vice President-Nuclear Operations.
J 4
i e
0 i
t Rev 2, 3/14/80
.41
- 18. NUREG.0578 Requirement 2.2.1.b"
Each licensee shall provide an on-shift technical advisor to the shift supervisor..The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shift technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.
Action To Be Taken at Big Rock Point Shif t. technical advisors (STA) having bachelor's ' degrees or equivalent and with specific training in plant response to off-normal events and in plant l
j accident analysis were on shift by January 1, 1980. These shift technical advisors will perform both the technical assistance and operational assessment functions specified in NUREG 0578.
The STAS who assumed duties January 1, 1980 have received seven weeks of training in plant _ design and layout, plant administrative and emergency procedures, response and analysis of the plant to transients and accidents and other topics necessary for successful commencement of watch standing duties.
They received one week of simulator training in February during which period they were replaced on shift by plant engineering staff members.
i l
Rev 2, 3/14/80
41a Candidates for shift technical advisor (STA) are required to have a bachelor's degree or equivalent in a scientific or engineering discipline (all current STAS are degreed engineers). Candidates for replacement STAS will be expected to connit to two years in this position. The training program for STAS includes the subjects listed below; nuclear plant experience may serve to reduce or replace certain aspects of this training provided the basis for such waiver is documented.
The training prograra for STAS includes:
1.
Plant General Employee Training 2.
Duties and Responsibilities of an STA 3.
Quality Assurance Indoctrination 4.
Plant Engineering Manual 5.
Radiation Work Permit Exemption and Respirator Qualification 6.
Basics of Nuclear Physics and Reactor Operating Response, Thermodynamics 7.
Plant Design and Layout (FHSR) 8.
Plant System Descriptions and Standard Operating Procedures 9.
Plant License and Technical Specifications 10.
Plant Administrative Procedures and Security Plan Rev 2, 3/14/80
41b 11.
Analyzed Accidents (FHSR) 12.
Transient Responses and Common Mode Failure Analysis
- 13. _ Plant Emergency and Off-Normal Procedures 14.
Health Physics Procedures: Sampling, Releases, Release Evaluations, Radwaste Transport 15.
Site Emergency Plan and Implementing Procedures 3
16.
Accident Assessment and Response 17.
Operating Experience Engineering Evaluation 18.
Simulator Training: Casualty Responses 19.
Management and Supervisory Skills 1
Requalification training for STAS will include the following material in addition to that required as part of General Employee Training:
-1.
Review of Facility Changes / Specification Field Changes i
2.
Review of Applicable Procedure Changes 3.
Review of Plant and Industry Licensee Event Reports 4.
Review of Health Physics Procedures for RWP Exemption 5.
Review of Revisions to Technical Specifications Tests 6.
Review of Changes to Engineering. Manual-Rev 2, 3/14/80
i 41c 7.
Review of. Revised Accident and/or Transient Analyses 8.
Simulator:
Casualty Responses 9.
Review of Technical Specifications Changes
- Training already completed for current STAS includes:
1.
Orientation 2.
RWP Exemption i
3.
Security 4.
Site Emergency Plan 5.
Technical Specifications 6.
' Operating Procedures
- 7. ' Final Hazards Summary Report 8.
Thermodynamics
.9.
Administrative Procedures 10.
DOT-NRC Waste Handling Requirements 11.
Health Physics Procedures - Inadvertent High Radiation Exposure
- 12.. Simulator: Casualty Responses i-Rev 2, 3/14/80 i
42 The major -function of the STAS is to perform an engineering evaluation of plant. operations from a safety point of view as outlined below:
1.
Operating history of the plant (equipment failures, design problems,
. operation errors, etc) and licensing event reports from other plants of similar design. The STA will provide any engineering assistance and recommendations to correct any of these problems on an as-needed basis or to avoid any reoccurrence.
2.
Plant ~ conditions required for maintenance and testing. Engineering judgment provided by the STA will be supplied, if needed, to assure that any testing or maintenance activity will not jeopardize the safety of the plant or specific piece of equipment being tested or worked on.
3.
The STA will review administrative procedures on an engineering evaluation basis as they apply to maintenance and testing of engineering safeguards equipment.
4.
Adequacy of plant emergency procedures. These procedures will be reviewed j
by the STA to identify potential problem areas from an engineering point of view. Upon implementation of emergency procedures, the STA will be ready to give engineering assistance that may be needed by the shift supervisor.
Consumers Power Company's (CPC) Topical Report No CPC-1A, " Consumers Power Company Quality Assurance Program Manual for Nuclear Power Plants,"-
establishes the requirements for QA. and technical evaluations for equipment Rev 2, 3/14/80
43 procurement and plant operation Quality Assurance. Therefore, this will not 4
be covered as an STA responsibility.
STAS will also review valve lineups as directed by the shif t supervisor.
The shift technical advisors will maintain a log in which complete entries on Items 1-3 above will be made as the actions occur. Additionally, the STAS will perform formal watch turnc iers and will use watch relief sheets which include the following information (minimum):
1.
Plant conditions at time of relief.
2.
Maintenance and testing in progress.
3.
Maintenance and testing planned.
The STAS report to an individual 'in a management position on the plant staff.
f This supervisor will present monthly updates to the Plant Review Committee on STA activities. The STA supervisor will also submit directly to the Vice President for Nuclear Op& rations a monthly memorandum on STA acitivites including a listing of the items accomplished under the operational assessment function.
4 Rev 2, 3/14/80
45
./-
Action To Be Taken Plant parameters are currently recorded on the control room yellow log sheet.
1The limits for the critical plant parameters are listed in the margin of the yellou log sheets. These limits are also included on a check sheet where the l
coming /offgoing shift supervisors and control operators document completion of i.r rev>.ew.
'b.
Assurance of the availability and proper alignment of all syc ems essential to the prevention and mitigation of operational transients and accidents by a check of the control console.
s
"(What to check and criteria for acceptable status shall be included on the checklist.)"
Action To Be Taken Switching and tagging orders, the caution tag logbook, the status board and 1'
the control console are checked to determine which equipment is either not
. ulable er operable. The equipment / systems that are not available or operable are listed in the shif t supervisor's log and control. room log and I
erified by both oncoming /offgoing shift.-supervisors and control operators at i
tha time of shift turnover.
]-
c.
Identification of systems and components that are in a degraded
- mode of operation permitted by the Technical Specifications.
For such systems and components, the length of time in the 4
degraded made shall be compared with the Technical Rev 2, 3/14/80 f
w
46 Specifications action statement.
(This shall be recorded as a separate entry on the checklist.)"
Action To Be Taken Where a degraded mode of operation is permitted by the Technical Specifications, the equipment / system in the degraded mode and the Technical Specifications action statement time criteria is recorded and signed by both oncoming /offgoing shift supervisors and control operators in the shif t supervisor's logbook and control room logbook at the time of shift turnover.
This action ident'ifies to the shift supervisor and control operator the amount of time the equipment / system can be left in the degraded mode per the Technical Specifications.
Item 2
" Checklists or logs shall be provided for completion by the offgoing and ongoing auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (w' at to check h
and criteria for acceptable status shall be included on the check _st)."
. Action To Be Taken The oncoming /offgoing auxiliary operators perform a review of the switching and tagging orders still open, the caution tag logbook, the control room console and the status board to determine which safety-related equipment is under maintenance or test conditions. Documentation of their review and a Rev 2, 3/14/80
\\
49 During accident conditions, the shif t supervisor remains in the control room at all times to direct the activities of the control room operators. He may be relieved by another qualified shift supervisor as directed by the
' Operations and Maintenance superintendent, or by the Operations supervisor. A proper watch relief must be effected prior to his leaving the control room.
I l
In certain rare instances, the shift supervisor may leave the control room during emergency situations. Prior to his leaving the control room, the shift l
supervisor will inform the No 1 control operator to assume command of the control room. The shift technical advisor remains in the control room during the shif t supervisor's absence to provide technical advise to the No 1 control l
- operator, if so requested. Administrative procedures will be revised to require the shif t supervisor to remain in the control room (unless relieved as -
above) whenever a site or general emergency is declared as specified in the BWR Owners Group positions submitted October 17, 1979 and approved November 14, 1979.
The Administrative Procedures, Chapter 4, and the Site Emergency Plan require the shift supervisor (or site emergency director if on site) to limit access into the control room during emergency or accident conditions to those personnel responsible for the direct plant operation and to those required to I
support plant operation during the emergency conditions.
l l
l Rev 2, 3/14/80 i
55 24.
ACRS Requirement: " Instrumentation To Monitor Containment Conditions'During the Course of an Accident."
Instrumentation shall be provided to monitor the following parameters after an accident:
1.
Containment Pressure. Measurement and indication capability shall extend from minus 5 psig to four times containment design pressure for steel containments.
2.
Hydrogen Concentration. Measurement capability shall be provided from 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
3.
Containment Water level. Capability shall include that level equivalent l
to 500,000 gallons for PWRs. For BWRs, levels from the bottom of the l
suppression pool to 5' above normal water level shall be included.
Instrumentation shall include continuous indication and recording and shall meet the design 'and qualification provisions of Regulatory Guide 1.97 including redundancy and testability.
l Action To Be Taken at Big Rock Point The required containment pressure monitoring capabilities will be provided.
l Instrumentation will meet the requirements of Regulatory Guide 1.97, Revision
'l (which references Regulatory Guide 1.89, Revision 0).
Containment pressure
~
monitoring capability meeting the specified range will be provided by January 1, 1981 in addition to existing instrumentation. Detailed design l
descriptions will be submitted for NRC information by July 1,1980.
Provision Rev 2, 3/14/80
l 56 of hydrogen monitoring capability will be deferred pending completion of a plant' risk assessment as' discussed in Consumers Power Company letter dated Feb rua ry 22, 1980.
Big Rock Point does not have a suppression pool. The requirement applicable to monitoring PWR containment water level is, therefore, more appropriate for
-Big-Rock Point than the requirement for BWRs. Big Rock Point's ECCS system is-designed to accomplish long-term cooling by removing water from the containment, passing it through a heat exchanger, and returning it via the
' core sprays. Switchover to this recirculating mode from the initial lineup involving direct injection of Lake Michigan water via low pressure core sprays is accomplished manually based on containment water level. Containment water
' level instrumentation, therefore, was a part of original plant design.
Continuous water level measurement exists from the level of the core spray.
1 l
recirculation pump suction strainers (574 feet) to a level approximately 6 to 9 feet above the maximum water level at which switchover.to recirculation mode
-is to occur (maximum indication at 596 feet compared to the lowest elevation inside containment of approximately 570 feet). This system will be modified to provide recording of the measurements by January 1,1981.
Additional containment water level indication is provided by four float-type level switches (elevations 574, 579, 587, and '595 feet) represented by console-mounted indicator lights in close proximity to the continuous indication readout.
A' redundant continuous measurement system meeting-the requirements fof Regulatory Guide.l.97,-Revision 1, will be installed. The redundant
~
--instrument reading will be recorded on a recorder separate from that to be used'for the existing instrumentation. The new redundant instrument will be
-installed by January 1, 1981.
Rev 2, 3/14/80
C-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 which, 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 s.ssumed source far exceeds radiation levels from any other source.
In fact, plant design is such that most systems which might contain radioactivity af ter 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 103 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 i
be essentially precluded by " containment shina."
Certain areas on site are sufficiently shielded that they would remain tenable following such a postulated accident. These areas principally are the control room, the shift supervisor's office adjacent to the control room (currently i
designated the interim Technical Support Center), and the area beneath the control room (currently designated the Operational Support 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 invento ry.
The radiation levels summarized above were calculated using source terms i
generated with the ORIGEN code. Locations of particular interest were evaluated using a virtual distance calculation method or by modeling the containment as an equivalent sphere with nuclide concentration adjusted by Rev 2, 3/14/80
- C-2 taking into account shielding afforded by structures within and outside the sphere. Site maps. showing approximate isodose lines for the first half hour are included herein.
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 (st!ch as manual switchover from injection to recirculation mode of core cooling) and actions which might be required in 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 Diesel 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 opera' tion under the above specified post-accident conditions
- indicates personnel radiation exposures above 25 Rem could be received performing these actions.
The back~y emergency diesel generator will be relocated to' an area near the primary diesel generator. Modifications will be.ade 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 af ter 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 failure 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.
Rev 2, 3/14/80
C-3 A new fuel supply tank will be i' 211ed. The new tank will provide a seven-day fuel supply.
Refueling af ter seven days under the above specified post-accident conditions should not result in personnel radiation exposures above 25 Rem.
Preliminary engineering for these modifications has begun. These modifications will be installed prior to January 1, 1981.
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 same procedures govern operation during nonaccident conditions. Precautionary 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 were in place prior to start-up from the early January 1980 outage.
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 the 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 was evaluated, for the long term, to enable personnel radiation exposures to be controlled to 10 CFR 20 levels un h t the specified conditions. The need for such shielding will be evaluated by nieans of a plant risk assessment as discussed in Consumers Power Company letter dated February 22, 1980. For reasons discussed at length below, this installation will be incorporated into modifications which may result from tha SEP if a shield is determined necessary. Actual installation of any additional shielding will, therefore, take place after the end of SEP.
j Consumers Power Company engaged a contractor, Catalytic Inc, to perform an evaluation 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 after the end of SEP, in order to permit engineering efforts to begin. These engineering efforts must proceed in parallel with SEP to generate the ~
information required to be integrated into SEP topic review.
(See V below.)
l The delay in initiating these engineering efforts which will result from Rev 2, 3/14/80
04 consideration under the plant risk assessment is not expected to hinder completion of either the SEP program or any shielding addition. The conceptual design review is complete.
Catalytic has recommended three options for consideration. These are l
(1) installing local shielding for specific plar.t areas, (2) erecting a concrete shield building a ound the existing containment, and (3) a combination of these methoos. Each of these options is discussed briefly below.
l The. conceptual design work performed by Catalytic was based on the following assumptions:
l l
l 1.
Core radionuclide release fractions specified in NUREG 0578.
2.
Instantaneous mixing of released radionuclides in the containment l
atmosphere (at time t = 0).
l 3.
Conservatively established criteria for post-shielding radiation levels in areas of concern.
l The conservative nature of these assumptions resulted in conceptual designs which are considered to be an upper bound on the modifications which will l
actually be required. Catalytic's analyses considered only the radionuclides mixed in the containment atmosphere since containment water level remains l
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 chielding which might be required for this option (SK-1300-1, Rev P) is included herein. This option includes addition of j
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 fiom che 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, l
l Catalytic's Option F consists of enclosing the existing containment within a shielding building of sufficient thickness to preclude the need for other l
shielding. The conceptual estimate (conservative) of the amount of shielding needed is a 56-inch thick coucrete 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 ahielding building is 36
- inches thick ano 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, shield thicknesses shown are probably above that which detailed design would show to be needed.
Rev 2, 3/14/80 m
~ - - -,
~
C-5 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 or more.
IV.
Qualification of Critical Equipment 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. Necessa ry i
equipment has been determined, based. upon information generated during review of various SEP topics. A tabulation of this equipuent is attached. The 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 plant design).
The radiation dose to which equipment outside containment might be subjected is likely to change with installation of additional shielding as discussed j
above. Review of this equipment to assure capability to withstand the expected dose will be accomplished soon after the final shield design is i
completed.
Equipment in the attached tabulation identified by an asterisk would not be affected by any of the proposed shield designs. The electrical equipment so indicated will be evaluated during an upcoming SEP effort on qualification of electrical equipment. The need to apply significant manpower to this SEP effort precludes work on the mechanical components until after the SEP submittal (approximately May 15, 1980). Evaluation of the mechanical j
equipment indicated by an asterisk will be complete by July 15, 1980. Efforts will be made to attempt to modify / replace any equipment found deficient in this review by 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 r
concern about radiation doses for equipment it protects.
a 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 i
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 1983.
In parallel with development of these criteria, existing plant structures are being modeled for seismic analyses. Any shielding added to these structurcs would require modeling and seismic analyses before it can be determined that the existing structuras, plus proposed modifications, are acceptable. The iterative procedure of design and analysis can begin only after the new site seismic criteria have
^
Rev 2, 3/14/80 i.
e r
.~,,
,a--.
+,
C-6 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.
i 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 I
assessment." Long-term safety would best be served by a detailed integration of shelding design and any modifications which might result from SEP.
VI.
Justification for Continued Operation The review described in II above has determined that the radiation exposures 4
to plant personnel can be limited to below life-threatening levels even in the unlikely event of an accident of the severity postulated in KUREG 0578. The high radiation levels which would be produced by such an accident would be localized and affect.the plant site only. No appreciable.ncrease in radiation exposure to any off-site individual would be avvected. 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 irdicate that continued plant operation pending integration of shielding design with SEP is justified.
1 l
1 Rev 2, 3/14/80
l l
i 1
r _
6 6
6 6
6 e
6 e) 0 0
0 0
0 t
0 C
vs 1
1 1
1, 1
a 1
Ol r
ia x
x x
x x
g xr eR e
e s(
1 1
1 1
1 t
1 vo o
n)
Dd 1
1 1
1 1
I s 1
e
( y d
dl o
o o
o o
a oe i
l ee t
t t
t t
6 t t l
0 a) 0 t e 5
5 0
1 aN 5
5 0
5 r s 8
r 0
0 0
0 0
3 0 gy ge 1
1 1
1 1
x 1
ea
/.
1 em r
t D 1
x x
x x
x 1
e xn
/
t inT
.v I 0 3
O 4(3 4
4 4
4 2
1 I
2 ve R
y N
rl O
ee I
t t T
l i A
an U.
i ei l
e rf I
A l.
ee V
e hd E
e T n N
i E
S e
d
.d l
O m
ey i e ol D
i l l T
ae a
i
~
N et re O
Si eS s
s s
s s
n P
I T
y y
y y
y ni n
A a
a a
a a
i f ti
)
l D
D D
l a e r a I
md om DA 0
0 0
0 0
en h e t
R 3
3 3
3 3
I SI TN E
D I
CCA er T
)
u SO s
P e
s 0
ss e
8 N
ea r
9 e
vh P
1 I
b l P D
n o
pa l
n E
o e
r uVf a
i R
i v
P k
e i
E t
l octi t
e d
D c
a e
t t a nl n
r e
I n
V v
n B ee e
u t
e y( VR r
s S
u l
e N
F t
a 6745V l
l e
kk s
l O
s V
9999 e
p nm f
cc e
p C
u 0000f n
p.eu f
- o. o. k m
r I I P
a t
4444 o a
uspu i
o E
h n
P S eO c D
. c c
B x
e VVVV n v
a l t u.
e E
V CCCC o d
nl oV r
en1 r
e i
e eat e
ne u
O b
T e
e oooot h
gV g
n nme s
r r
t t t t a S
o s n i
opp o
o T
e e
l rt ni a
sia l
c c
t N
h h
rrrro r
t na r r
r tp eeu t
e s
n p
p i iii s i
i Q
'9 S
S AAAAl A
NVHD S
PI E E
es l
o I
U d
Q E
de t
g c
n ep i
z x
i e
l a
r i
o p
t s
f n
E e
n m
l 1
o p
e 4,56 e
i n
4 1 1 1 t
t j
888 a*
a 4567 t
23456 a
1 ig 99 99 555555 P
5 777 i s 4567 c
4 ce 6666 i
E 0000 1 1 1 1 1 1 4444 999999 6
- ** ov 6666 f
2 V
AhC sl i
I l i VVVV VVVVVV C
Ooo I iI s a SSSS l
CCCC SSSSSS C
P PPP TTTAV PPPP au Q
2 3
4 5
6 7
i i
J l
C-2 Integrated Dose Over E.gn ipmen t -
Function Time Needed Time Neeled (R.s.ls) 8.
PS-636
- t10-7068 Timer Sta rt Containment Spray 4 x 105 to 1.1 x 106 PS-637 *
!!O-7068 Timer Start Delay Needed Only 20 tlinutes.
(Integrated Over PS-7064A
- HO-7064 Timer Start Thereaf ter itemain Scaled 30 Days)
PS-706411
- 110-7064 Timer Start Iniletinitely.
9.
tio-7067
- Turbine liypass Isolation 30 Days 1x 105 to 4 x 105 CV-4106
- Bypass Warmup Control Valves Hust SV-4916
- Air to CV-4106 Stay Closed CV-4104
- Steam to Seals and Air Ejector SV-4899
- Air to CV-4104 CV-4200*
Turl>ine Stop Valve CV-4117
- Reactor and Fuel Pit Drain luolation SV-4922*
Air to CV-4117 CV-4102
- Clean Enclosure Sump Isolation SV-41195
- Air to CV-4102 CV-4103*
Dirty Enclosu're Sump 1 solation SV-4896*
Air to CV-4103 CV-4105*
1)cmin Water Isolation SV-4897*
Ai r to CV-4105 10.
Elect ric Fire Pump ECCS 30 Days 1x 105 liiesel Fire Pump 0
11.
PS-612 Diesel Fire Pump Start 2-5 llours 2x 10 PS-615 Elect ric Fi re Pump Start
- Qualification for expected dose to be completed in 1980.
Rev 2, 3/1h/80
/,
(**
C3 Integrated llose Over Equip..nt-Functit..
Time Needed Time Needed (Itads) 12.
SV-4919 Fuel to Diesel Fire Puup 30 Days j
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 Low Fire System Pressure PS-789 Through PS-796 Fire Pump Discharge Pressure 5
6 13.
PT-173
- Reactor Buildinn Pressure 30 Days 4 x 10 to 1.1 x 10 PT-174
- PT-18 7
- 5 14.
HUS Channel 30 Days 1 x 10 A-D Sensor and Actuator Cabinets 3
15.
125 V 11-C llCC 30 Days 1 x 10
- 1 and #2 Transformers 5
16.
DOI 125 V D-C HCC 30 Days 1 x 10 D02, plo -
125 V D-C Distribution Panet D03 Station Battery Charger Switchgear Station llattery, 4
17.
'llPS A-D lininterrupt ible Power Supply 30 Days 3 x 10 4
18 P2A*
Core Spray Pumps 30 Days 3 x 10 P211*
0 4
19.
PS-638*
Core Spray Pump Discharge Pressure 30 llays 1.5 x 10 to 3 x 10 DO-7066*
Post incident System Valves
!!O-7072*
20.
Diesel Generator, 30 Days 1x 102 Auxi l i a ries,
Control Panel 6
21.
Elect rical Penetrations
- 30 Days 1.1 x 10 oqualification for expected dose to be completed in 1980-Rev 2, 3/1h/80 i
4
" BIG ROCK POINT -
PREDICTED RADIATION LEVELS AT t = o HOURS FOLLOWING AN ACCIDENT WITH FISSION PRODUCT FRACTIONS OF 100% NOBLE GASES AND 25%
HALOGENS IN CONTAINMENT ATMOSPHERE"
-7 N
p
\\
L
\\
s s
\\
\\
\\'/\\
~.
t s
\\
i
,.,x o
\\
/ / /\\
l[
\\
s ll" s
/
s/
N g
g N
4 g
/
s s
\\
/
o
'N
. s o
4~
T**
\\
\\
0 vi 's 0
'J h,
Q I
_ u..}-.. " '
00
/
- ' r j-c/
m oo 3o
- C y
7 o
I E
/
1 3,000 h-E S0*
g
\\
H s
O t
I M
X u
N q : 57 y'
- p'.
Rev-2,-3/-lh/80
_ _. _.. L
" BIG RCCK POINT -
PREDICTED RADIATION LEVELS AT t = o HOURS FOLLOWING AN ACCIDENT WITH FISSION PRODUCT FRACTIONS OF 100% NOBLE GASES AND 25%
HALOGENS IN CONTAINEIT ATMOSPHERE"
~
o s
7
\\
\\
s s
N N
\\
\\
.k f ~ *. '..
- s.
V:}:.u.[d s'-**.
i-
~
g s
m-g o
0 o
- O
~4 c'
n e
L'
/
k.
'.Y/l... s -
r-I l/
Nl oo go
_e Q*00 f
/
[
]
i 5'000_, -
o, y x
)
=
.ro /
I E
i.
8 g
c/
1,00,
)
~
0 w
/
/
5
/
/
F#
l
.2 i
e
- 3. x o 3:.
l l
eco g.
1,7 i
f
(( = GIS
~
T Rev 2, 3/1h/80
APPENDIX G Primary Containment Isolation System Data
[
I
u ga l
iJi
,d g
jl
\\ Ml
\\q\\ \\
4 3lU E
9
\\\\ \\'\\ \\ \\ \\\\ \\ \\\\ \\ \\ \\ \\ \\
?
\\l\\
t,l C
O '2 2
N I
i m
2 *!
}.
6 l
l cI d0 5
w l
l t
t ze da 1; TQ X XlX IMA
/iXI XI / AlkkXIM OX m
d'3
" = al ES: 33lB i 8l 8
@ e l @ / l l /' l8 Ele g 8 g! g 2
$3 El1
'~3 El Sls 8
8 9310 $J 34 813 0 8F%@
w oI oIo o
o o ! o yo u:olololo o oq o o
- i g "'"A ll o I-a1o o1 o
o ( o ro ul u ojo o!o un o o
ij ml oyoIo' o
o ojo 0 010 0 0 0 0 0 o 0
5 "Ni oIIoloi o o: olo o ulo o o o c ol o o
j]Em as! sol A 3 el B o l XIX f8 A&p) ygente~j ps gfjTif A
si
. !H j 3 M IA @l3 M Q 4
IJ h ij
_,If]!wr1 e
e kkh 3; !kF !!$
- v% uLju$hN 3 %J er to #s c n 934 4'R41 10 3"il511 3 si i !!!! i si!!! !Ii!B ! 1 j edEl a4i e4iae ia 34 l l Iledii :a !seis<iet b e4 e< f [ $0.Qi s1 8.i X M %i l 1 M I ', ; %l% S S!8t%iX:3 d / S i ~~ j{ ! !n$bi.me#!! ss' d w w a I =. s al s 1
- s 1
1 %l a 1 5 (pp a 4 o,,O!ill : 3 3 6 !2 3: 2 1, 6 13 c y alai. gt la M'Q h), f ~=cmd3: J A A J t s 1 t
- t i
j i i t .c s a 1 F4 / I j 1 A q ) i 2 9
- m ai xi i
m 1 1 m v m M g g 404 3, 3I li 5 53 'l i 3! ' i 3 33: 3 !5 (EM:'
- i.-a n. wad
- i. l. l!.!
\\' l i. f B i.] ! y O r S c. e .,h 5 nu c a p h 3! M h! b bl h b. I h hM3 3 1 l 8 57 G M M lll,atinssii2 212 a al 31 y s!i g 2:12 315!!'alais 3 3 y ic $ 3Y# sis: 515tX Xla 3 ails N JfX Fl 3 Ji X 313!! Xill X 3 0 r 5 r;n s10 3a 0! $1I til fislostilal s 1 li t tit 1 1 0 $ f,i 54EE @e$ @ lil$6 <S i $! d F3@ l5 6$ t! "dO m-3 o b 4 4 Mr-'r e a at ra < e~ v +a aa >s t s +amza sa m s v a g 3 a h ,x,'JC L, o! 4 oloI w
- u. Ie!
u I e I s3" M W! F WlI' !l!! E il 3 Sl! d 'ai%'A 8 E Wlai 8 0Idi 8 2 3 -WMia 2 i Jt3' 2 2 12 i s 3 3 "WP) 3 3 5!4i i i l 3,4 A i 39f L >g ie !E I ~ hh d$ kd g$ge g u2! $ $c g 5> 150 d d E s$ od Q u6 mm 3 sg o be i y3ske y i s ~ on -m _1 s sieye< a l-i ei m. 9
de l 8 y 5 i5l \\ \\ \\ \\ \\\\ \\\\ \\ \\ \\ \\\\ \\ \\ \\ \\ \\ \\ 2 H2h e ihj l4!j .N i \\ \\ \\ 7 3 l 3 I E i w 2 h b 4 1 %X X)(/.XX]X X XK X'.XA X gisiglg al L If , <A gg g xAg
- xils s alg xts gG x. xl y >ix l
O h "numi Wl9 @ X. glg E G3@ 3;l@ 3 1DllB 318 81 @ @l Klx! vl TG;td o a ollo o clot o clu o J! U U)o OQO o A1 ) i
- <3"i@i o o o ojol o o o! o giv v vi o ulo uno W 'lill sib y lj nres o o o ol olo o o! o o!lv v v. o olv ut al o! tiAi /. l A
/ l me o oo alo o o o! o oju o ou o! U 0001 ol [ [ ar !ge:gs xlx!i.x!x!!Mug simpgsix1xb ti., 1-y!=enmXxjp[$l!!}$el=8dillIl}}h!!M =g p llilll is g l}# HI i e i um i l m a s eist e l 3 5%%'l > 'is#s4} 44 li \\ /l se een i sd I \\ / 'li. i 'l s j lj g'#0?M i I1si$lil % \\ !i \\ /I M Ml !!i %l \\/ 'l\\ V i $h c3 /\\ \\l $ .\\- 4 \\ \\. l O <6-J, l s.. }#*}P: e\\l 9 / \\ = mun 5
- E r
,h '\\ o fm<C / sj y q. \\ a e y# a a \\ a . = 3 5 I 1s15 'I il IJ 23 la' 12 2; (iJl N N j smuj' ~ \\ z! r [,{ j,/ {! I li s / j W l,o p g el o g r il s li w% x 4,f h SNE1l ' l 3 I Ul \\l l S $l \\/ I i i il \\ Ul bS 0 0 %l5 3 -33G li (s l ili g g! g o t 'I 5 sip 3 5l A l. 1 ll '1 1 g g g M gg \\ p / d s s e es, r TQ$ $ M / \\ il ? m W mi \\ l i. ? / al N cf s A g 20#M 21x! 2 3i 5! 5 3!2!! a al2 z!2121212 si 1121 f $i } D'2 3 J; 5 fi si a 3 ltix. lis $15: 313; ! aiX Ji t / li tle E29 2 m so mx zi i 32 2 mo o s e .'r E=9 c9e. l ra, m-e 3.. o8 g yJ g, s g o e e s s s 6 a e o o 8 N 3/G'1/\\ 3 3 I T S b N-A 7 h i d h 3 3 ss BB BB 4 M9 9 33 3 3,3 3 g, 2 2 2, z i cc, nWWe x I a 4xi z 1 o I o. I e X X MSRI GM lalII i G F I IB G E D E!!lfi cf 13'Ce I M I d W 5iE E E E Ofl3 DT: smit a I 2 i a
- s
- n3nx 1
8 96 2l1 "?f;s ' i i ! l 1 3Iisa ! I sl ill! bfb gg ) fa dh fl ]$f y]2 DdMM OdM{ n g d i n h l ef o + in j g l t-l gl e g g g vN Wul . -.. ~ O
[ .df wl o 35-ss b
- b"$
l l C E as m 6 3d d I S [ I S M, h W\\ \\ g 3 A E 1, IXalX@BsslXllgigh *~X g ~ l tmevj xlx 'oJiX IXxlN ^!7j x X J w 9 3(X 8H Xi@ XjX XWl/ X 6 g % 41o1010110 o o dio o MAti
- 24; 010 l W 110 0 0 U Q O O if till'/; \\!l\\/
I p op5 0 0 oilO O O olO O \\ \\ ( l' ! ilh l T^5NIO OudoO Ool0Ol l, T l ',l/ tj J M D )< X!3WX!X!Xix_lXLX)/ lx}ixiXu El!=mlW )e$\\/ s! wlele W i w ee z: 2 2 lj $0 aq p a 2 glfjjjj j j y j j gjjj g 2 g l 4i t d,.X4&j i El - / 1 j,3%"W \\ isd N / \\ / l, 6 It i \\ l Q M I\\/lXl \\/ \\/ LDjil i; q g '\\ [\\j S!!\\ GE55 [ h wa%! !\\ h /\\ a I n=mi , sl / 9 \\ e N4 T a i 3 u b i; s f $N^?;). l\\ i l it i) M s a
- =J \\/ 1/
\\/ U M M p 7 7 MWi \\/ l \\/ \\/ yl lIl y UlU C j 6 ??al A /\\ /\\ ti 3 4ti C-f l 3 (5'2E' \\ j\\ l. l!\\ c\\ r\\ h p erem / \\
- 3
( ) 1 2 .l wnA / \\ M,, (< i) g 8 34 g Of&% 2l2 21212 213121 a ,h Ji, st (8t? [,' j M4:1 JIX MtXIX Xix13tX !! AI r. \\- u 8 2 y^ m o anu nuo n, r v [{ l7Y f d $ 6 l ennl h e e o e w 9 6 3 3 2 g @ a > e t-t jt t l. 2 = 4 =s e s saA >r EYacYi,., d l 4 F .XiXIX Xs 5 se ei n e m a u 'O'.7,';'4 9! al E E Sif s s \\! %TM 1 \\ g 2 3l2\\ s z t! -w! ai a m a i i r, x v N k43 U { G04q og on s t b o $ c$ 0 G, 0 2 1BBe 6 8 u iN e il s! gr a $19: $1s R i D
a f 6 4 8 e 4 2 3 d5 4 h i l=i 1 g s i s s 8 a n a s g ? 8 b 5 hE 5 d i I h h, k hNYs* a s i hs @ e 5-3 E S e a g 3 oob o 2 R O g j g b 0 b 1 O b } I] O ! i I.f 0 3 a q$ u! pg f .o s q j i i i l 4 g i z }a: H a 1 I si I j U. 0 1 8 ! j s% 8 o a l l 1 v 3 3 e a Ug t, s e E J $y 0 $ h b h 9 m 3 a e o a 2 b b b b d bhs3bb $ i
- g6 i s
u, s n i p o s p-s n 2 g f 2 8 3 e y yha a g z 0$ 6 6 ' g } Y E 5 bb b !8 9 t f'"*f ^j )s a a
- s 8 3 i J
E ));! f f $ $ 3 ip i n jo s 4 9 0 3 4 8 y 1 9e olgl 1 k 5 $k 3 hg $ 3 a l i d I ha i }p$3 p-j! 3 j i s a; 3 . o n ns s m s e 2 t f ie l ? a i l j 'f ,3s } J gd j j# g l d i e 3a j jE i ^d e sj 3g es s a a r s w e 1 5 4 8 0 t $ 8 s t e e l sc us e, o s o g.a S 4 ej 'J 9 J f ]"j 4 $ J J J J a3 a $ 3 3 4 4 j la!$)e$li ho l 3 2 e g a ! j d3 0 !! j 3 E I d b n# 4 "19di i u b ns o
- 1 Hi q!!) 3 2 :
a y i s 1 i s ! j fS iis jg D 1 e e d2 3 3 6 3 3 I d s l a n g lj g 8 g23 }a$lD $ $5 i ll lilllll} } } } l l l l $33i 1$} } ~.~--,o.au
Y *$ c.(. g ? i.">.7r Di/142RM S CF L~oM ~n9tNM.CNT /.foLA7"/C// ( $ <? f 2 / t~ E S v rn I:: xm
- %\\
' tf ns A v a r.c a w O' c.v . M rn I ^ FJ6 8 /.!*,y7~ Ex &A.us r \\ I lc fII 4 1 D t 5t / l L ?:*:5 / i=/4 C V w r Et% '4~ y cpc ~ ~ V 9* \\ 5:Wbws' L/dr M DPC ~ Ds:c:wwie r:D/.We Pen-522J' 72"$ 7f.'> 3r I F/e D Q.C w '5wrP Pt.er? y I .Ciscs.nac,4* LJ Rev 2, 3/1h/80
/.%-@ 77 i Diswer$ c.c" 6 vrnsWM.!~M7~ ACtAr/CW 2"h ( Cwwura 5 .? E FM E osart.sexp Puan p< X l osscana u vic-3of bene.v wa=se.s l v4-i t fic;. f tu:r.euurxr Au ,\\'n> (. 'w VA-7 na
- s. nae sn x
N2 VA.2cc AJ.L*7 ,. y M 4 tO" tasso M F/6 H c c m Morra y ,& tru22p m.m/Cdu Nw/ r.r, rsa. Q M ar noa a /xt:.evn J fo-cmco / CH/Ca*.'ssrsh. & Cd.C*k nur sera 7:o r pu.22 so::.ca. D~ x + x4... sx_. JL ._2 Rev 2, 3/ik/80-
P. "I k } ' % s*= '7h ( 7 0/Ad%A,W3 CJ" $0Na*4/MMfsV7" /:bt.s97/OW ZAs. erg"RZ [ [.a e w.r.s ) ev c i eia 7 m,a Sr.rw L>uis ~< 7 Mo-70 65 EU.cTkscAt.t.Y D/ssa.aaj {Vsuf I -y l ru, x n acra m s ~, 4N l 1 vy vs. 1 vrw.9 204 W o 301 Y gS 7"*P W vRD-34 [oxTrcs Ai!:C M ' l.- HrnasesLsc S V r:? x ff d/ l Pa w ' ><- i van..to 4 w vr p.- ' Pas;- /m sosW;-i v? < tot. ~ //d' Y j h pes Suns *t. y e co. l ,1, i L@ + al m..i I ,,A, .A. 7/e N ,&seu.Ler g Q; e ?' Wfs ~ I L c -us v-u c. w a s, %:swas um=to ~ _s v .tewori c.(~/acwwm.w" w w,.A =m. - ~.... - wo. j i D m7 D
- D ~T' I
nev 2, 3/1h/80 .J w J . 2. a
>. ;+r-,r ' &- $> o$ Q. . p,x.sms os nraine.vr /:.=vrica B'wca.s (C.erwu o ) cv
- M fic-o 7,eenre.o Ma:re
'l '/, >c seruna cucu v-w nsraa rnauu zuur a,- ruwer. c+ to.:1 mrrs r,ici .oza,a e,ac. ~ b i Fie,- P h.:acrcz //iset. Ar M Stara m urp. a VP l*3Cl, ~ ~ " Elb & Fosi~ hC/Dev7~ EJcxus l i o. i Wil5
- l..O.
.R COR.E 6PRSf HG f NA /el /N RMP Dsc.ut.M': VPI-Loo VPI-4 @X h L.C_. L.0-L.o. VPL-l VFL-2 VPI-?, AG. E C o n n S -9.e< k up~RETp9M hl L -4 dYPI-3 L.O. D" D
- D ~T I
9 Rev 2, 3/1h/80 C J WJ .Q. .:3 ,}}