ML20235Y659
| ML20235Y659 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/16/1987 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20235Y661 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96 SAIC-87-3056, TAC-51270, TAC-51271, NUDOCS 8710200440 | |
| Download: ML20235Y659 (10) | |
Text
a r J SAIC-87/3056
. TECHNICAL. EVALUATION REPORT FOR WISCONSIN ELECTRIC POWER COMPANY'S POINT BEACH NUCLEAR PLANT, UNITS'1.AND.2:
SAFETY PARAMETER DISPLAY SYSTEM SAFETY' ANALYSIS REPORT TAC NO. 51270/71' October 16, 1987 l
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Scence Appicatons kwernationa/ Cc,poratan d
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- TABLE OF; CONTENTS-
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Section
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.-INTRODUCTION.......'.,.......
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II.
SUMMARY
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EVALUATION
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Description:
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Parameter' Selection.....,..........,.
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' Display Data Validation....
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D.-
Human Factors Progrart....
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IV.
CONCLUSION................-..=......-
6 REFERENCES....c.
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j TTECHNICAL EVALUATION REPORT l
FOR
- j WISCONSIN ELECTRIC POWER COMPANY'S POINT BEACH NUCLEAR PLANT, UNITS 1-AND.2
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SAFETY PARAMETER DISPLAY SYSTEM
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I.
INTRODUCTION All holders of operating licenses issued by the; Nuclear. Regulatory.
Commission (NRC) and applicants for an operatino license' must provide 'a J
Safety Parameter Display. System (SPDS) in'the cytrol room of their-plant.
j The. commission approved requirements for the SPDS.are defined in NUREG-0737, Supplement 1 (Reference 1).
The purpose of the SPDS is to provide a concise display of critical 3
plant variables to control' room operators to aid-them in ' rapidly and reliably determining the safety status of the plant. ~ NUREG-0737, Supplement 1,
requires licensees and applicants to prepare a written safety.. analysis' describing the basis on which the. selected parameters are sufficient to assess the safety status of each identified function for. a wide range of
- events, which include symptoms of severe ~ accidents.
Licensee's and j
contains schedules for design, development, installation,'and full operation applicants must al so prepare an implementation plan for the SPDS which j
of the SPDS as well as a design verification and validation plan.
The safety analysis and the implementation plan are in be submitted to the NRC for staff review.
The results from the staff's review are to be published in a Safety Evaluation Report (SER).
i Prompt implementation of the SPDS in operating reactors is a design goal of prime importance.
The staff's ' review of SPDS documentation for operating reactors called for in NUREG-0737, Supplement 1 is designed to
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avoid delays resulting from the tima required for NRC staff review. The NRC staff does not review operating reactor SPDS designs for compliance with.the requirements of NUREG-0737, Supplement 1 prior to implementation.unless a pre-implementation review has been specifically requested.by licensees. The licensee's Safety Analysis and SPDS Implementation Plan is reviewed by the NRC staff.only to determine if a serious safety question is posed or if ths 1
analysis is seriously inadequate.
The NRC staff review to accomplish-this I
is directed at (a) confirming the adequacy of the parameters selected to l
provide information about critical safety functions (CSFs), (b) confirming that means are provided to assure that the data displayed are valid, (c) confirming that the licensee has committed to a human factors program to ensure that the displayed information can be readily perceived-and' comprehended so as not to mislead the operator, and (d) confirming that' 'the f
SPDS will be suitably isolated from electrical and electronic interference with equipment and sensors that are used in safety systems.
If, based on this review, the staff identifies a serious safety question or seriously inadequate analysis, the Director of the Office of Enforcement or the j
Director of the Office of Nuclear Reactor Regulation may require or dire' t j
c the licensee to cease implementation.
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.I II.
SUMMARY
Science Applications International Corporation (SAIC),
as technical assistance contractor to the NRC, has reviewed the Point Beach Nuclear Plant, Units 1 and 2 SPDS Safety Analysis Report (SAR) with respect to the j
issues noted above, except for electrical isolation (which is being reviewed separately).
First the review team concluded that the parameters selected for SPDS did not include Residual Heat Removal (RHR) flow or its equivilant,
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stack radiation monitoring, containment isolation indication, and j
containment hydrogen concentration.
Second, the licensee did not provide
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information regarding the verification and validation program used to ensure valid display data.
Third, the licensee did not provide information regarding a human factors program.
Based on an evaluation of the information submitted to NRC, the review team concluded that continued implementation of the SPDS wouldmot present a se'rious safety concer,n.
Ill. EVALUATION By letter dated June 30, 1986 (Reference 2) the Wisconsin Electric Power Company (WEPC0), in response to NUREG-0737, Supplement 1 submitted a Safety Analysis Report (SAR) regarding the Safety Parameter Display System for the Point Beach Plant. This Technical Evaluation Report discusses SAIC's review of the SAR and presents its results and conclusions regarding the proposed SPDS.
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' A' j$PDSDescription;
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i As.' described in" the SAR, thel Point B' ach:SPDSL is being"developedias? ' a,
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e subs' t. ' of' the Safety Assessment. System (SAS). ~ The. SAS :is designed-to?
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' provide the following capabilities's:.
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Display of. plant mode dependent' key parameters" that are ' associated.
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-with safety status of;the plant; W
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ProvideL thirty-minuteL trend graphs for groups-of: ' rel Ate'd) h a
parameters.
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Providei cues to operators aboutlthe likelihood offthreel' major; p
accidents (Loss of. Coolant ~ Accident (LOCA)', Steam' Generator. Tube
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Rupture (SGTR), and Loss of Secon'dary Coolant.(LOSC)),. using ian' Accident Identification' and Display System _ (AIDS).-
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. Provide the operator with a Critical. Safety Function Monitor.which determines the conditions to. assess the status.of.the six. Point U
Beach SPDS critical safety functions.
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.I Sional Validation d
The licensee 'did not provide information on t'e means to: assure.-that h
the display data would be valid.
J SPDS Availability 1
The licensee stated, the ShDS will: be available' du~ ring normai plant h
operation.
In this context, design availability is understood to : encompass j
the following minimal functional capabilities:
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The ability to monitor and display' the status of-all CSFs' inL'at'..
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- l least one location in the control room.
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The' ability.to determine the value of all; variables:which are used in the CSF status determination in at'least one locationiin the
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q SPDS Use and Location-i1 The SPDS displays.are located in the controll room, -the Technical-L Support Center,.and the computer room.
No information was-provided^on the-
'I accessibi.lity of the displays to operators in the vicinity o f.' ' t'he main!
' control board, f
Modes of Ooeration The. Point Beach.SPDS is designed.to operate during normal plant-i
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operation as well as during and following 'an accident.
N Data Storaae The Plant Process Computer' System (PPCS) is capable o'f storing all 1
inputs to the SAS.
The data are ' stored on hard' disk at: five-second intervals for forty-eight hours, and at ten-minute intervals for two. weeks.
B.
Parameter Selection 1
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Section 4.1.f of.NUREG-0737, Supplement I states thati "The minimum information to be provided shall be sufficient'to provide-information to plant operators about:
(i)
' Reactivity Control (ii)
Reactor core cooling and heat "Omval from.the primary-system 1i (iii)
Reactor coolant system int @ U j
(iv)
Radioactivity' control (v)
Containment conditions" k
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For review purposes, these five item 2 have been considered as Critical, Safety Functions.
'In its review, the NRC staff has considered the Westinghouse Owner's.
j Group Emergency Response Guidelines (ERGS) Program, which was reviewed and 4
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approved by the Staff (Reference Sj,{ U = fincipal technical' source.'of, l
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'WEP'0$ sad'itsselectionofparameters'on parameters.important to safety.
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the ERGS, Revision _
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which were also used to ' develop plant ~. specific:
j Emergency Operating Procedures and Critical Safety functisu Status Trees for j
L Point Beach Nuclear Plant.
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A~ review of-the Point Beach SARLindicates that the SPDS parameters selected-' satisfy most of - the flVREG-0737, Supplement 1 critical -safety functions.
However,'the Point Beach SPDS does not include all parameters suggested by NRC ' for parameter selection. The parameters that-are.not provided include:
residual. heat removal ~(RHR) system flow or equivalent'
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indicator of,RHR heat removal,' stack radiation ~ monitoring, containment isolation indications, and containment hydrogen concentration.
i RHR flow is 'a key parameter for determining the validity of' heat removal when the secondary system is not the principal means.of.'. removing.
decay heat (i.e., large LOCA or normal-shutdown RHR). ~It4 is recommended.
L that the licensee either add this parameter or other parameters indicating j
the same function to the SPDS-l a
q For pressurized' water reactors, direct release of radiation to.the
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atmosphere can occur through_ stacks and the main steam safety valves.
The j
i stack monitors are normally used to' measure. fission products that may be.
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released during normal operation, or during an accident if the containment
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is not isolated. Since this is a key parameter for. monitoring 'a ' direct radiation release path to the' environment, it <is recommended that the p
licensee include this parameter in the SPDS design or justify not. including
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l Containment isolation is recommended as a key indicator of. containment conditions.
Although not explicitly tied to the ' operator decisions by the
- ERGS, failure of containment isolation can lead to direct release of radioactive materials to the atmosphere. By monitoring the status of all l
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isolation valves, there is assurance that known pathways penetrating the containment have been secured.
'It is' therefore recommended that.the licensee either add containment isolation to the. SPDS' or justify not l
including it.
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Containment hydrogen concentration is~a key parameter :for. monitoring.
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containment combustible gas control.
Depending on the type 'of containment oxygen concentration is also required to be monitored.
Since.this parameter is essential for ' determining the containment structure integrity during_
certain accident scenarios, it is recommended that the licensee either i
includes containment hydrogen concentration in the:SPDS design or justify not including it.
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l In ' conclusion, we -determined that the-propo' sed Critical;. Safety j
Functions and. associated parameters for,the' Point Beach SPDS comprise Lan j
appropriate' set with;a few exceptions. These include indications for heat-l removal when steam generators are isolated, stack. radiation monitoring,
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containment isolation valves status, and containment hydrogen concentration, l
The licensee should~ add these parameters to 'the SPDS or justify not including thec.
1 C.
Display Data Vslidation The licensee did not provide information on activities that should be conducted to assure that the displayed data are valid.
I D.
Human Factors Program According to NUREG.-0737, Supplement 1 (Section 4.1.e); the SPDS shall j
be designed incorporating accepted Human Factors Engineering' principles.
In its SAR, the licensee did not provide information.regarding a-human factors program for the.SPDS.
IV.
CONCLUSION The review team conclusions are provided below:
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The parameters selected for display.are consistent with the. NRC
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staff-approved Westinghouse Emergency Response Guidelines, but do-not include some parameters necessary for critical safety function 1
assessment. The system did not. include RHR flow, stack radiation monitoring, containment isolation status, and containment hydrogen concentration.
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o The licensee did not provide information on the verification and validation of.SPDS display data.
o The licensee did not provide information indicating.the existence l
of a human factors program for the SPDS.
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REFERENCESi 1-i.
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l 17 NUREG-0737,JSupplement-1 Requirements ~for Emergency Response Capability-
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i(Genericcletter 82-33), December ~17,71982.
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Letter, C W.
Fay, ; Wisconsin Electric.Po'werLCompany,(to G. lL6ar,: NRC.,1 d
1 June 30;.l1986,;with: attachment.
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Safety l Evaluation'of " Emergency Response, Guidelines, R Generic" Letter, 1
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't 63-22,.Ju.ne'8, 1983.
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