ML20114B633
| ML20114B633 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/23/1985 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8501290382 | |
| Download: ML20114B633 (14) | |
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DUKE POWER GOMPANY P.O. BOX 33180 CitAHLOTTE, N.C. 28242 HALH. TUCKER TE1. EPITOME vws enemanzar (704) 373-4531 January 23, 1985
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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50--414
Dear Mr. Denton:
-Ms. E. G. Adensam's letter of December 18, 1984 transmitted a request for additional information regarding the Catawba Nuclear Station's Safety Parameter Display System.
The attached response addresses the specific concerns of the staff regarding the SPDS.
Very truly yours,
/$
Hal B. Tucker WLH: sib Attachment cc:
Mr. James P. O'Reilly, Regional Administrator NRC Resident Inspector U. S. Nuclear Regulatory Commission Catawba Nuclear Station Region;II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Robert Guild, Esq.
P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 21351s Devines Street Columbia, South Carolina 29205 Mr. Jesse L. Riley Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207
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DUKE POWER COMPANY CATAWBA NUCLEAR STATION RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION CATAWBA SAFETY PARAMETER DISPLAY SYSTEM January 22, 1985 r
I.
THE APPLICANT'S SUBMITTAL, H. B. - TUCKER (DUKE) TO H. R. DENTON (NRC), MARCH 28, 1984, ATTACHMENT SECTION 4, DOES NOT PROVIDE EYIDENCE THAT CRITICAL SAFETY FUNCTIONS FOR THE SPDS, SECTION 4.1(f) of NUREG-0737, SUPPLEMENT 1,
WERE CONSIDERED IN THE SELECTION OF THE CATAWBA SPDS VARIABLES.
A.
PROVIDE A LISTING OF THE VARIABLES PROPOSED FOR THE DISPLAY ON THE CATAWBA SPDS THAT CONSTITUTE SUFFICIENT INFOIMATION TO ALLOW CONTROL R0(M OPERATORS TO ASSESS THE STATUS OF:
1.
REACTIVITY CONTROL 2.
REACTOR CORE COOLING AND HEAT REMOVAL FRCH THE PRIMARY SYSTEM 3
REACTOR COOLANT SYSTEM INTEGRITY 4.
RADIOACTIVITY CONTROL I
I 5.
CONTAINMENT CONDITIONS
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B.
SHOW HOW TPE VARIABLES PROPOSED FOR THE CATAWBA SPDS SATISFY THE MONITORING REQUIREMENTS FOR EACH OF fHE FIVE CRITICAL SAFETY FUNCTIONS SPECIFIED IN NUREG-0737, SUPPLEMENT 1.
RESPQNSE:
I Section 4.1.4 of H.
B. Tuoker's (Duke) To H. R. Denton (NBC),
March 28, 1984 submittal contains six (6) Critical Safety Funo-tions- (CSF) which, for Westinghouse plants, address the same functions referenced in section 4.1(f) of NUREG-0T37, Supplement 1.
The basis for the selection of these six CSF's is contained in the document " Background Information for Westinghouse Owners -
Group Emergency Response Guidelines", HF/LP-Rev. 1, dated Septem-l bec 1, 1983, under the tab named " status Trees".
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7 DUKE POWER COMPANY CATAWBA NUCLEAR STATION SPDS RESPONSES January 22, 1985 Page 2 Described below are the six Critical Safety Functions as defined for the Catawba units.
These.six CSF's correspond to the five proposed in NUREG 0737, Supplement 1 as follows:
1.
" REACTIVITY CONTROL" is addressed by the Critical Safety Function, h haritianH tv.
The variables monitored by the Catawba SPDS and the basis for thsir selection are described
_ in section 3 23 1, pages 3-94 and 3-95 of the attachment "A",
excerpted from Duke Power Company, Catawba Nuclear Station, Emergency Procedure Guidelines Reference, Volume I, dated June 1984.
2.
" REACTOR CORE COOLING AND HEAT REMOVAL FR(M THE PRIMARY SYSTEM" is addressed by two CSF's:
Egr.1 Cooling described in section 3 23.2; and Heat Aink described in. section 3.23 3. See pages 3-95, 3-%, and 3-97 of Attachment "A".
3
" REACTOR COOLANT SYSTEM INTEGRITY" is addressed by two CSF's: Integrity described in section 3 23.4 beginning on page 3-97; and Inventory described in section 3 23.6 on pages 3-100 and 3-101, 4.
" RADIOACTIVITY CONTROL" is addressed by the Containment CSF described in section 3 23 5 on pages 3-99 and 3-100.
5.
" CONTAINMENT CONDITIONS
- is addressed by the Containment CSF described in section 3 23 5 on pages 3-99 and 3-100.
i II.' ' DESCRIBE THE PROGRAM FOR VALIDATION 0F THE CATAWBA ' SPDS VARI-ABLES.
IN THIS DISCUSSION DESCRIBE HOW THE PLANT SI)RILATOR, i
AND/OR CONTROL' ROOM WALKTHROUGHS OF SPECIFIC TRANSIENTS AND ACCIDENTS WILL BR USED TO DEMONSTRATE USEABILITY OF THE SPDS l
COVERING INSTRUMENT SETPOINTS FOR SYSTEMS ACTUATIONS AND OPERATOR I
ACTIONS.
RESPONSE
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The Catawba SPDS is based upon the six (6) Status Trees, one for.
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each CSF, as defined in the Westinghouse Owner's Group Emergency Response Guidelines and specifically developed for and applied to the Catawba Units in Duke Power's " Emergency Procedure Guidelines for' Catawba." The acceptability of Duke's implementation of the Westinghouse Owners Group. Emergency Response Guidelines is documented :in Section 13.5 2 of " Safety Evaluation Report related to the operation of Catawba Nuclear Station, Units 1 and 2' dated December 1984.
DUKE POWER COMPANY-
- CATAWBA NUCLEAR STATION SPDS RESPONSES
' January 22, 1985 Page 3 L
Catawba's Status Tree based SPDS takes credit for the validation programs which were performed on the above documents, and specif-ioally, those associated with the definition and selection of the six CSF's and associated variables which are required to deter-aine the safety status of the plant.
Further, the usefulness of the Status Tree based Catawba SPDS 4
system was confirmed in a. task analysis performed by Duke Power's 1
Control Room Design Review Team.
The SPDS and supporting dis-plays were reviewed by the Control Room Review Team which oon-tained control room operators, design engineers, and a human factors engineering consultant.
A review and evaluation of the SPDS display system was performed
- to ensure that the system provides direct, readily usable infor-mation organized in an effective format to support operator task requirements.
The human factors review was conducted in two separate _ activities:
(1) a task analysis conducted during the display system development and (2) a human factors survey of the laplemented displays.
i-Task Analysis The task analysis activity of the SPDS human factors evaluation was conducted using the control board mockup which had been fabricated for use in the task analysis activity of the Control Room Design Review (CRDR). An event scenario was developed using the plant emergency procedures and the Westinghouse - Emergency Response Guidelines. The scenario provided an ordered framework of a set of possible responses to an initiating event against~
which the system was evaluated.
From the event scenario, plant parameter inputs to the SPDS logio were identified.
Values for these parameters were developed consistent with plant conditions for several selected tino intervals during the duration of the scenario.
Selected time intervals were chosen to be one minute intervals from initiation until 5 minutes after initiation, and 10 minute intervals from 10 minutes into the scenario until 30 minutes after initiation.
The SPDS logic output states were determined for each time interval using the specific plant parameter values. PhotograValo slides were then produced for each time interval to represent how each SPDS and supporting display would appear for that time interval.
1
' DUKE POWER COMPANY-CATAWBA NUCLEAR STATION SPDS RESPONSES-January 22, 1985 Page 4 A walk-through of -the event scenario was performed by a task analysis team consisting. of a senior reactor operator and a mechanical / nuclear systems engineer.
During the walk-through, the operator performed the task actions required while the
. engineer served as observer.
In addition, several other members of the Control Room Review Team served as observers and slide ooordinators.
The proper time sequenced slides 'for the SPDS display were projected onto the SPDS display CRT mockup to simulate the action of this display for. any of the secondary supporting displays at a particular time interval were projected onto the supporting display CRT mockup in response to operator command, simulating the call-up feature of the supporting display system.
The usability and effectiveness of the displays were evaluated by the task analysis team using a set of pre-selected task analysis principles.
These principles covered such items as logical ordering of displays, terminology and abbreviations, labeling, coding, usability of displayed information, and operator task support.
In general, the task analysis activity evaluated the SPDS. and supporting displays to determine if the displays pro-vided a. logical, readily usable format to support the following operator tasks:
- Monitor Critical Safety Function Status (CSF)
- Observe CSF status changes
- Determine which CSF is degraded
- Determine severity of degradation
- Identify component / functional area out-of-tolerance
- Determine which confirming displays and restoration procedures to use
- Monitor restoration process
- Monitor remaining CSF status during restoration Human Factors Survey A human factors survey of the actual SPDS and secondary support-ing displays as implemented on the control room CRT displays were performed.
During the survey the control room CRT displays and the operator keyboard were used to call-up, observe, and review
. each separate display.
In addition, the displays were reviewed
.during a simulated alarm condition.
1
' DUKE POWER COMPANY CATAWBA NUCLEAR STATION-SPDS RESPONSES January 22, 1985 Page 5 The survey evaluated the format and arrangement of the displays and the operator keyboard interface using applicable survey
- principles from the Control Room Survey Principles Checklist
'which was derived from NUREG-0700 for use in the CRDR. These
. principles covered areas such as color, usage, character height, room lighting and glare, presentation of data, labels and coding, operator message presentation, and the arrangement and uce of the operator keyboard interface.
.Results The results from both the task analysis and the human factors survey were documented in the form of recommendations for design changes to the SPDS and secondary suppceting displays.
These recommendations concerned items such as audible alarming upon a change of CSF status, the addition of CSF status blocks to the bottom of the supporting displays in addition - to those on the primary SPDS display, alam message format, display function button position on the operator keyboard, and double spacing _of lists for readability.
The human factors recommendations from each review activity were resolved and the required changes to the SPDS display system were implemented.
In summary the human factors review activities determined that the. SPDS and supporting displays and the. operator interface provided readily usable and easiJy comprehended information in an effdotive format to support operat e task requirements.
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ATTACICIENT A 3 23 F-0 Critinal Rafety Funntion Status Trees 3 23 1 F-0.1 Suboriticality
.h aator Trin Reauired The Suboriticality CSF is designed to monitor the post-trip reactor status.- This branch point results in a GREEN status during normal power operation.
Power Ranae 461 Following a reactor trip, nuclear power promptly drops to only a few percent of nominal, and then decays away to a level some 8 dcoades less.
Decay heat levels resulting from radioactive fission product decay are never more than a few percent of nominal power and also decrease in time.
Safeguards heat removal systems are sized to remove only decay heat and not significant oore power. The 55 level was chosen because it is~ olearly readable on the power range meters.
Nuclear power above,5%, in a core that is supposed to be shutdown, is considered an extreme challenge to the fuel clad / matrix barrier and a RED priority is warranted. The appropriate guideline for function restoration is FR-S.1, RESPONSE TO NUCLEAR 20WER GENERATION /ATWS.
Inter==Ainte Banae SUR Zero or Maantive At this point, power range flux has been determined to be not significant, so no extreme challenge exists. However, a positive startup rate (SUR) in the intermediate range will shortly lead to power production if operator action is not taken,.since no inherent feedback mechanisms exist below the point of adding heat. A positive SUR is considered a severe challenge to the CSF and an ORANGE priority is warranted.
The appropriate guideline for function response is ~ FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /
ATWS.
" Source Range Energized This decision point is used to determine if further evaluation should be directed at the source range flux behavior, or back at the intermediate range channel indications.
Inter-adiate Danae SUR More Haantive Than -0.2 DPM l
Normally, following reactor trip, intermediate range flux decays at a toonstant -0 3 DPM. A rate of decrease less negative than -0.2 DPH (e.g.,
-0.1 DPM) is considered to represent an unsatisfactory condition and a YELLOW priority is warranted.
The appropriate guideline for function restoration is FR-S.2, RESPONSE TO LOSS OF CORE SHUTDOWN.
If the rate of decrease is less negative than -0.2 DPM, then the CSF is satisfied.
Source Range SUR Zero or Negative Normally,. following reactor ~ trip, neutron flux decreases into the source range and staya there. Typically source range count rate fluctuates, and does not exhibit any sustained increasing trend.
Such a trend, as indi-3-94
cated by a positive SUR, is considered an unsatisfactory condition and a YELLOW priority is warranted.
The appropriate guideline for function restoration is FR-S.2, RESPONSE TO LOSS OF CORE SHUTDOWN. If source range SUR is zero or negative the CSP is satisfied.
3 23 2 F-0.2 Core Cooling Core Rwit TCs (1200*F Analyses of inadequate core cooling scenarios show that core exit temper-ature greater than 1200 F is a satisfactory criterion for basing extreme operator action.
The average of the 5 highest oore exit 5 thermocouples should be reading greater than 12000 F.
Five has been chosen to allow margin for individual thermocouples failing high.
This temperature indicates that most liquid inventory has already been removed from the NC system and that oore decay heat is superheating steam in the core.
An extreme challenge. to the fuel matrix / clad barrier is imminent and a RED
' priority is warranted. The appropriate guideline for functional response
'is FR-C.1, RESPONSE TO INADEQUATE CORE COOLING.
NC System Anhanolin, >0*F If NC system subcooling is less than O' F, then SI flow should be maintained to the NC system to provide inventory makeup and the Core Cooling CSF is not satisfied.
Subsequent blocks check for inadequate or degraded oore cooling conditions.
If greater than O' F NC system subcooling is indicated, then the CSF is satisfied.
At Laast one NC Pn=a 1hinni n,
~The'RVLIS design has two ranges relevant for core cooling, lower range and dynamic head range, for use without NC pumps running. and with NC pumps running, respectively.
This block determines which range should be used
- to assess the Core Cooling CSF status in subsequent blocks.
If any NC
-pamp is running, then the RVLIS dynamic head range should be used in assessing core cooling conditions.
If no NC pump is running, then the lower range should be used.
Core Rwit TCs 4700*F l
~ If.the average of the 5 highest oore exit thermocouples indicates greater l
than 700 F, superheat at the core exit is indicated.
An inadequate core cooling condition will exist if, in the next block, RVLIS indicates less than 4455 (6 feet) collapsed liquid ' level in the core.
If core exit i
' thermocouples indicate less that 700'F, then an inadequate core cooling
~ condition does not ' exist and the subsequent RVLIS check will assess whether a degraded core cooling condition has been reached.
BYLTR Lower Ranea >4M (Core Rwit Ta-ner atures Greater than 700'F) 1 If RVLIS lower range is less than 435, then the core is uncovered and an inadequate core cooling condition has been reached.
A RED priority is warranted and FR-C.1, RESPONSE TO INADEQUATE CORE COOLING, is the appro-priate guideline for functional response. If RVLIS lower range is greater than 435, then a degraded oore cooling condition exists since the core 3-95
exit temperatures are greater than 700' F from the previous block.
An ORANGE priority is warranted and FR-C.2, RESPONSE TO DEGRADED CORE COOL-ING, is the appropriate guideline for functional response.
RYLTR Lower Ranaa >4tt (Core Rrit T==neratures Lens than 700' F)
If RYLIS lower range is less than 435, then the core is uncovered, but since core exit temperature has not reached 700* F, an inadequate core cooling condition has not been reached. A degraded core cooling condition exists. An ORANGE priority is warranted and FR-C.2, RESPONSE TO DEGRADED CORE COOLING, is the appropriate guideline for functional response.
If RVLIS lower range is greater than 45%, then only a saturated core cooling condition exists.
A YELLOW priority is warranted and FR-C.3, RESPONSE TO SATURATED CORE COOLING, is the appropriate guideline for functional response.
RYLTR Dvnm=in Hand Ranma > Satooint If an NC pump is operating, then even under a highly voided NC systen condition the core exit thermocouples can be expected to indicate satu-rated temperatures.
This block checks for NC system voiding less than approximately 25 percent which, if NC pumps are subsequently stopped, would ensure the core would initially be kept covered and adequately cooled 6 If RVLIS dynamic head range is less than the indicated setpoint a degraded core cooling condition exists.
An OPANGE priority is warranted and FR-C.2, RESPONSE TO DEGRADED CORE COOLING, is the appropriate guide-line for functional response.
If RVLIS dynamic head range is greater than the indicated setpoint only a saturated core cooling condition exists.
A YELLOW priority is warranted and FR-C.3, RESPONSE TO SATURATED CORE COOLING, is the appropriate guideline for functional response.
3 23 3 F-0.3 Heat Sink wancow Ranae Laval in at Lammt One SG > 51 (>181 ACC)
A level in the narrow range in any steam generator, including a ruptured one is sufficient to ensure an adequate secondary inventory for a secon-dary heat sink. If level is not in the narrow range, the operation of the feedwater systems will determine whether a loss of secondary heat sink is imminent.
Anviliary Faaduatar Flow to SGs > 450 ann Total auxiliary feedwater flow of greater than 450 spa ensures that, in the absence of narrow range level in any steam generator, the capability of auxiliary feedwater to restore level and maintain a secondary heat sink is available.
If not, then an extreme challenge the heat sink CSF is imminent and a RED priority is warranted.
The appropriate guideline for functional response is FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.
Pr===ure in All SGa 41210 nmia In the event that pressure in any steam generator is greater than the highest steam line safety valve setpoint, then the steam generator design limit may be _ exceeded and integrity may be challenged.
Also, there is no.
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flow path in use removing energy from that steam generator. The Heat Sink CSF is not satisfied and a YELLOW priority is warranted. The appropriate guideline for functional response is FR-H.2, RESPONSE TO STEAM GENERATOR OVERPRESSURE.
unprow Ranee Level in All SGs (82.41 (C67.41 ACC)
An overfeed due to excess feed flow or a steam generator tube rupture may lead to a high level in a steam generator.
This block checks all steam generators to ensure identification since this condition may cause un-wanted atmospheric releases or challenge steam generator integrity.
Note
. that although the level in the affected steam generator may reach the top of the narrow range span, significant volume still exists before the steam generator fills with water.
The Heat Sink CSF is. not satisfied and a YELLOW priority is warranted.
The appropriate guideline for functional response is FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL.
Pr===nce in All SGa 41175 nmie If any steam generator safety valve is open, then an unisolable heat removal path is being used.
A better path is to use steam dump to con-denser or SM PORVs which are controllable and isolable.
Also, condenser steam dump will not release steam to the atmosphere. The Heat Sink CSF is not satisfied and a YELLOW priority is warranted. The appropriate guide-line for functional response is FR-!I.4, RESPONSE TO LOSS 0F NOHMAL STEAM RELEASE CAPABILITIES.
unprow Ranee Level in All SGs >51 (>181 ACC)
Feedwater should be maintained until all steam generators are in the narrow range unless a faulted steam generator is identified. Narrow range level is reestablished in all steam generators to maintain symmetrio cooling of the NC system. If any level is low, the Heat Sink CSF is not satisfied and a YELLOW priority is warranted.
The appropriate guideline for functional response is FR-H.5, RESPONSE TO STEAM GENERATOR LOW LEVEL.
3 23.4 F-0.4 Integrity T==nerature Deoramme in All Cold T=== (100*F in f mat 60 Minutes If the temperature decrease in any cold leg has exceeded 100*F in the previous 60 minutes, then there is a potential concern for thermal shock.
If not, then no other checks on rate-dependent limits are necessary. The remaining conceras are NC system overpressure and cold overpressure which will be checked in subsequent blocks.
If the temperature decrease has exceeded 100*F in the previous 60 minutes, the degree of cooldown must be i
assessed before a thermal shook concern can be identified.
This is j
checked in subsequent blocks.
All NC Svaten Pressure-Cold Les T==nerature Points to Rieht of f.imit A l_
The objective of Limit A is to provide a limit that indicates an extreme l
thermal shock ooadition. The basis of this limit is to prevent growth of a flaw in the vessel._ If Limit A has been exceeded, then operator action is necessary to limit further NC system temperature decreases or NC system 3-97
pressure increases.
A RED priority is warranted since an extreme chal-lenge to the CSF is occurring and FR-P.1, RESPONSE TO IMMINENT PRESSURIZED-THERMAL SHOCK CONDITION, is the appropriate guideline for functional response.
All MC System Cold La= Tannaraturam 1950*F If any cold leg temperature is less than 350'F, then operator action is necessary to minimize further NC system temperature decreases and NC system pressure increases. An ORANGE priority is warranted since a severe challenge to the function exists and FR-P.1, RESPONSE TO IMMINENT PRES-SURIZED THERMAL SHOCK CONDITION, is the appropriate guideline for funo-
-tional response.
Pr==="v'fum>+ Pressure <2400 nmia (2250 nmia ACC)
Since pressuriser pressure should normally decrease following an sooident, this setpoint is not expected to be exceeded except for pressurization transients such as spurious safety injection or a power generation heat removal mismatch.
The pressurizer PORY lift setpoint is 2335 psig, so a pressure of 2400 indicates PORY malfunction, or other inability to handle the transient, and therefore a possible challenge to the pressurizer code safety valves.
All NC Svaten Cold Lea Ta=narature >457'F The temperature region between 457'F and 350*F is intended to allow time for operator action to try to prevent entering a region of potential thermal shook.
In this region the CSF is not completely satisfied and a YELLOW priority is warranted.
The appropriate guideline for functional response is FR-P.2, RESPONSE TO ANTICIPATED PRESSURIZED ' THERMAL SHOCK CONDITION. If all NC system cold leg temperatures are greater than 457'F, then the Integrity CSF is satisfied.
Pr===urimer Pr===ure (2400 nmfa (2260 nata ACC)
Since pressurizer pressure should normally decrease following an aooident,
.this setpoint is not expected to be exceeded except for pressurization l
transients such as spurious safety injection or a power generation heat l
removal mismatch.
The pressurizer PORY lift setpoint is 2335 psig, so a pressure of 2400 indicates PORY malfunction, or other inability to handle the transient, and therefore a possible challenge to the pressurizer code
' safety valves.
1 NC Svaten T==narature 7100*F In order to determine if cold overpressure is a concern, a check is made on whether NC system temperature has decreased to below the temperature at which the pressurizer PORV low setpoint should be enabled.
Subsequent i
j blocks check if a cold overpressure condition exists.
L NC Svaten Pressure 4400 nat a l
If the. pressurizer PORY low setpoint should be enabled and NC system pressure exceeds 400 psig, then action may be necessary to minimize or l
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decrease NC system pressure. The priority of action will be determined in subsequent blocks.
If NC system pressure has not exceeded the cold overpressure limit, then the Integrity CSF is satisfied.
All NC Svaten Cold Lea Ta-naratures > 250*F If cold leg temperature in any NC system cold leg is less than 250*F and NC system pressure is greater than 400 psig, then a severe challenge to the function exists and operator action is necessary to limit NC system pressure.
An ORANGE priority is warranted and FR-P.1, RESPONSE TO IMMI-NENT PRESSURIZED THERMAL SHOCK CONDITION, is the appropriate guideline for functional response.
If all NC system cold leg temperatures are greater than 250*F, then even though the cold overpressure limit has been exceeded (previous block),
there is no extreme or severe challenge to vessel integrity, even at very high pressure. A YELLOW priority is warranted, however, since the CSF is
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not satisfied and FR-P.2, RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK CONDITION, is the appropriate guideline for functional response.
3 23 5 F-0.5 containment Cantain==nt Pr===ure (15 nmi-If containment pressure is greater than design pressure, an extreme challenge to the containment 4 arrier exists.
The challenge does not necessarily come from the pressure alone, but rather from the potential pressure spike which could result from a hydrogen ignition.
/dso, above containment design pressure, leakage may exceed design basis limits.
It is expected that containment pressure suppression equipment should-be able to maintain pressure below design pressure.
If not, then operator action is necessary to check containment functions and a RED priority is war-ranted.
The appropriate guideline for function restoration is FR-Z.1, l
RESPONSE TO HIGH CONTAINMENT PRESSURE.
Contai nment Prammure d't omie Pressure above 3 psig indicates s significant energy release to contain-ment and merits prompt operator action to ensure operation of containment i
pressure suppression equipment and performance of Phase B isolation. Such a pressure also requires Steam Line Isolation and is considered a severe challenge to the containment barrier and an ORANGE priority is warranted.
The appropriate guideline for function restoration is FR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE.
containment Hydronen Conaantration <0.51 Appreciable accumulation of hydrogen gas inside containment is not ex-peoted except for inadequato core cooling scenarios.
When significant' s
amounts of hydrogen are generated by the metal-watse reaction in the core, the gas may be released into the containment ats, where more quickly than the electric hydrogen recombners can remove it.
In this case it is important to have an anticipatory hydrogen concentration setpoint to provide maximum opportunity for the various sitigation systems to reduce the concentration before it reaches flammability limits which would cause l
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a containment pressure integrity concern due to a hydrogen burn.
Containment han Level (19 ft.
High energy line breaks could result in a large volume of water being pumped into containment. As the water level rises, it might threaten the availability of equipment required for long term cooling of the core and/or containment.
Such a high water level is considered a severe challenge to the containment barrier and an ORANGE priority is warranted.
The appropriate guideline for function restoration is FR-Z.2, RESPONSE TO CONTAINMENT FLOODING.
ContainmantRfdiationMonitors(9R/hr i
Normally, containment building radiation levels are, fairly low and con-stant.
However, during an accident, significant radioactivity may be released into the containment atmosphere.
In-containment systems are available to filter and scrub the contaminants from the atmosphere, and radiation alone does not repressat a threat to containment integrity.
This is considered an unsatisfied condition and a YELLOW priority is warranted. The appropriate guideline for function restoration is FR-Z.3, RESPONSE TO HIGH CONTAINMENT RADIATION. If containment radiation monitors indicate less than 3R/hr, then the CSF is satisfied.
3 23.6 F-0.6 Inventory r
Pressurizer Level (Q21 (<801 ACC)
This decision point allows proper resolution of the actual inventory condition in subsequent decision blocks.
If pressurizer level is above the normal operating range, the next decision block determines if it is due to excess inventory or voids in the vessel.
If level is not high, then further questions check for low level and voids in the vessel.
RVLIS UR Indicates >071 and Stable (Pressurizer Level > Q21 (7801 ACC)
Having already determined that pressurizer level is high, this question tries to define the cause.
If the upper head region is full, then the problem is simply one of excess inventory; the Inventory CSF is considerod not satisfied and a YELLOW priority is warranted. The appropriate guide-line for function restoration is FR-I.1, RESPONSE TO HIGH PRESSURIZER LEVEL.
If the RVLIS does indicate' voids in the upper head region, then j'
the problem is likely due to some type of bubble in that region.
Since the presence of a bubble is considered an unsatisfied condition, a YELLOW priority is warranted. The appropriate guideline for function restoration y
FR-I.3, RESPONSE TO VOIDS IN REACTOR VESSEL.
is Pressurizer Level >171 (>Mi ACC)
This block is entered after having determined that pressurizer level is not high.
If level is also not low, then the pressurizer inventory is p
considered satisfactory and a further question is asked about reactor vessel level.
If pressurizer level is not greater than the indicated
~#
setpoint, then the problem is one of low inventory, with or without voids in the veasel. The condition is considered an unsatisfied condition and a 3-100 i
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YELLOW priority is warranted.
The Core Cooling Status Tree checks for more severe or extreme challenges to Inventory that also challenge the Core Cooling CSF.
The appropriate guideline for function restoration is FR-I.2, RESPONSE TO LOW PRESSURIZER LEVEL.
HYI.TM UH Tndicates > 071 and Stable (Pressurizar Level Determined to be Nomal)
Having determined that pressurizer level is normal, the remaining inven-tory question relates to water level in the reactor vessel. If level does not indicate that the vessel is full, then some type of voids are present in the vessel upper head.
The presence of a bubble is considered an unsatisfied condition and a YELLOW priority is warranted. The appropriate guideline for function restoration is FR-I.3, RESPONSE TO VOIDS IN THE REACTOR YESSEL.
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