ML20011E782
| ML20011E782 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/08/1990 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20011E781 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 9002220464 | |
| Download: ML20011E782 (4) | |
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UNITED STATES l'
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ENCLOSURE 2 SUPPLEMENTAL SAFETY EVALUATION
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l TENNESSEE VAlt.EY AUTHORITY l
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296 r
CONFORMANCE TO REGULATORY GUIDE 1.97 1
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1.0 INTRODUCTION
The staff completed its review of the the Tennessee Valley Authority's (TVA or the licensee) confomance to Regulatory Guide (R.G.) 1.97, Revision 3, by providing the staff's Safety Evaluation to the licensee on June 23, 1988.
The staff found that the licensee's design for the Browns Ferry facility was acceptable with respect to conformance to R.G.1.97 with the exception of the following variables: core s residual heat removal (RHR) pray flow, low pressure coolant injection flow, system flow, RHR heat exchanger outlet temperature.
l cooling water temperature to ESF system components, cooling water flow to ESF system components, emergency ventilation damper position and neutron flux.
By letter dated August 23, 1988, the licensee requested that the staff reevaluate these issues.
The issues involved with core spray flow, low pressure coolant injection flow, RHR system flow, and emergency ventilation damper position were subsequently resolved by the staff in its response to the licensee dated January 19, 1989.
A detailed review and technical evaluation of the RHR heat exchanger outlet temperature, cooling water temperature to ESF system components, cooling water flow to ESF system components, and neutron flux issues was perfomed by EG&G Idaho, Inc., under a contract to the NRC, with general supervision by the NRC staff.
This work was reported by EG&G in Technical Evaluation Report (TER),
"Confomance to Regulatory Guide 1.97: Browns Ferry-1/-2/-3", dated August 1989
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(attached).
We have reviewed this report and concur with the conclusion that the licensee either conforms to, or has adequately justified deviations from, the guidance of R.G.1.97 with the exception of the variable neutron flux.
2.0 EVALUATION 1
We have reviewed the evaluation performed by EG&G contained in the attached TER and concur with its bases and findings.
We agree with the EG8G findings that the licensee either conforms to, or has provided an acceptable justification for deviations from, the guidance of R.G.1.97 for the following variables:
a) RHR heat exchanges outlet temperature, b) cooling water temperature to ESF system components, and c) cooling water flow to ESF system components.
We also agree with the EG&G findings that the licensee does not conform to, or has not provided an acceptable justification for deviations from, the guidance of R.G.
1.97 for the following variable: d) neutron flux.
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a)
R. G.1.97 recomends Category 2 RHR heat exchanger outlet temperature instrumentation to monitor the operation of the RHR system.
The licensee has provided instrumentation that meets the Category 2 criteria except for j
environmental qualification.
The licensee has also provided torus (suppressionpool) temperature,suppressionpoollevel,drywell temperature, drywell pressure, reactor level, and reactor coolant system pressure instrumentation.
This instrumentation as a minimum meets the Category 2 criteria of the regulatory guide.
The licensee states that the torus temperature, when trended, can be used to monitor the performance of the RHR heat exchangers.
The emergency operating instructions (E01s) use i
the remaining instrumentation as primary indicators for determining the heat energy remaining in the containment.
The staff finds the licensee's alternate instrumentation and associated EDIs acceptable for monitoring the RHR heat exchanger outlet temperature.
b)
R.G.1.97 recomends Category 2 cooling water temperatgre to ESF system n
components instrumentation with a range of 40 F to 200 F to monitor the operation of the cooling water system.
The licensee has provided i
instrumentation that meets the Category 2 criteria except for environmental qualification of a cable that passes through the Unit 3 reactor building.
The display for this instrumentation is logated in ghe Unit 2 control room.
The range of this instrumentation is 90 F to 100 F.
The licensee has committed to provide environmental qualification for the cable passing through the Unit 3 reactor building, bringing the entire instrument loop up to Category 2 requirements.
The staff finds this acceptable, The licensee states that river water is the source of cooling water for all ESF components.
The river temperature can only change slowly over a long period of time.
Therefore the operators of any unit have timely access to the temperature information in the Unit 2 control room.
The staff finds this deviation from the control room display recomendations of the regulatory guide acceptable.
0 Thelicenseejustifies.thesmalltgmperaturerange(90Fto100F)by stating that temperatures below 90 F are not of concern as the EgF components have a maximum design cooling water temperature of 95 F.
The licenseealsostatesthattherive5temperatureandthecoolingtower discharge channel do not exceed 95 F during plant operation.
The staff finds the range of the cooling water temperature to ESF system components instrumentation acceptable.
c)
R.G.1.97 recomends Category 2 cooling water flow to ESF system components instrumentation to monitor the operation of the cooling water system.
The licensee has provided instrumentation that meets the Category 2 criteria except for environmental qualification.
The licensee has separatedthisinstrumentationintotheRHRservicewater(RHRSW)andthe emergencyequipmentcoolingwater(EECW).
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The licensee has committed to provide environmentally qualified instrumentation for the RHRSW instrument loops.
This is in accordance with the recomendation of the regulatory guide and, therefore, the proposed upgraded instrumentation for monitoring the RHRSW flow is acceptable.
The EECW system is shared between the three units. There are four pumps, each powered by diesel generator-backed panelboards.
Each pump has circuit breaker status lights, pump motor ampere indication, and flow indication available in each control room.
This instrumentation is located in a mild environment. The licensee states that only two pumps are needed for full capacity of the EECW system.
The licensee also states that either of the two redundant systems can provide the complete cooling of the connected ESF heat loads because each component is supplied cooling waterfromboth(viacheckvalves)EECWheaders. The EECW system valves are administrative 1y controlled in a preset position that apportions cooling water to the ESF components.
Based on the licensee's description of the EECW system and associated instrumentation, the staff finds this alternate instrumentation acceptable, d)
R.G.1.97 recomends Category 1 neutron flux monitoring instrumentation to monitor reactivity control.
The licensee has provided neutron flux monitoring instrumentation which complies with the Category 1 criteria except for source and intermediate range monitor drive mechanisms and controls and the neutron monitoring system power sources.
The staff's position, provided in its June 23, 1988 Safety Evaluation, is that the existing neutron flux instrumentation is acce atable for interim operation only and that the licensee shall install and lave operational neutron flux monitoring instrumentation which fully confonns to the recomendations of R.G. 1.97, Revision 3.
The licensee has stated that their current position on neutron flux monitoring is consistent with the Boiling Water Reactor Owners Group (BWROG) generic position documented in NED0-31558, " Position on NRC Regulatory Guide 1.97, Revision 3, Requirements for Post-Accident Neutron Monitoring System".
The licensee also stated that they would conduct a plant specific evaluation of the BWROG's position pending NRC acceptance of this generic position.
By letter dated February 29, 1990Property "Letter" (as page type) with input value "05000250/LER-1979-020, Has Been CancelledUnable to interpret the "February 29, 1990" input value as valid date or time component with "Month 2 in year 1990 did not have 29 days in this calendar model." being reported." contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., the staff rejected the BWROG position.
It is the staff's position that the licensee should evaluate the newly developed neutron flux monitoring systems and install neutron flux moni-i toring instrumentation which complies with the Category 1 criteria, of Regulatory Guide 1.97.
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4 3.0 _ CONCLUSION Based on the staff's review of the enclosed TER and the licensee's submittals, we find that the Browns Ferry Nuclear Plant, Unit Nos.1, 2 and 3 design is acceptable with respect to confomance to R.G.1.97, Revision 3, with the exception of the neutron flux variable.
The staff finds acceptable the existing neutron flux instrumentation for interim operation.
It is the staff's position that the licensee shall install neutron flux monitoring instrumentation which complies with the Category 1 criteria, of R.G. 1.97, Revision 3.
An appropraite implementation schedule will be developed by the Project Manager via discussion with the licensee.
Once the schedule is established, the licensee is required to inform the Comission, in writing, of any significant changes in the established completion schedule identified in the staff's safety evaluation and when the action has actually been completed.
Principal Contributor:
B. Marcus Dated:
February 8, 1990 t
o-gg EGG-MA-6873 Revision 1 TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97:
BROWNS FERRY-1/-2/-3 Docket Nos. 50-259/50-260/50 296 Alan C. Udy Published August 1989 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 TAC Nos. 51073/51074/51075
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SUMMARY
This EG&G Idaho, Inc., report documents the review of the Regulatory Guide 1.97, Revision 3, submittals for the Browns Ferry Nuclear Plant, Unit Nos. 1, 2, and 3, and identifies areas of nonconformance to the regulatory guide.
Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.
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FIN No. A6483 B&R No. 20 19-10 11-3 Docket Nos. 50 259/50 260/50 296 TAC Nos. 51073/51074/51075 11
i PREFACE This report is supplied as part of the " Program for Evaluating Licensee / Applicant Conformance to RG 1.97," being conducted for the U.S.
Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, Division of Engineering and System Technology, by EG&G Idaho, Inc.,
i Regulatory and Technical Assistance Unit.
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CONTENTS j
SUMMARY
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PREFACE...............................................................
iii 1.
INTRODUCTION.....................................................
1 2.
REVlfW REQUIREMENTS..............................................
2 3.
EVALUATION.......................................................
4 3.1 Adherence to Regul atory Guide 1.97.........................
4 3.2 Type A V a r i a bl e s...........................................
4 3.3 Exceptions to Regulatory Guide 1.97........................
5 4.
CONCLUSIONS......................................................
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5.
REFERENCES.......................................................
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_,3 CONFORMANCE TO REGULATORY GUIDE 1.97:
BROWNS FERRY 1/-2/ 3 1.
INTRODUCTION
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On December 17, 1982, Generic tetter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability.
These requirements have been published as Supplement No. I to NUREG 0737, "TMI Action Plan Requirements" (Reference 3).
Tennessee Valley Authority, the licensee for the Browns Ferry Nuclear Plant, provided a response to Section 6.2 of the generic letter on April 30,1984 (Reference 4). Additional infomation and corrections were submitted on May 7, 1985 (Reference 5). Additional information, justification, and schedules were submitted on August 23, 1988 (Reference 6).
These submittals address Revision 3 of Regulatory Guide 1.97 (Reference 7).
This report is based on the recomendations of Regulatory Guide 1.97, Revision 3, and compares the instrumentation proposed by the licensee's submittals with these recommendations.
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2.
REVIEW REQUIREMENTS s
1 l.
L Section 6.2 of NUREG-0737, Supplement No.1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies with Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that l
provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97, l
1.
instrument range t
2.
environmental qualification 3.
seismic qualification l
4.
quality assurance 5.
redundance and sensor location 6.
power supply 7.
location of display 8.
schedule of installation or upgrade l
The submittals should identify any deviations taken from the regulatory guide recommendations and provide supporting justification or alternatives for the deviations identified.
l Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March,1983, to answer licensee and applicant-questions and concerns regarding the NRC policy on this subject.
At these meetings, it was noted that the NRC review would address only exceptions taken to Regulatory Guide 1.97.
It was also noted that when licensees or applicants explicitly state that instrument systems conform to the regulatory guide, no further staff review would be necessary. Therefore, 2
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this report addresses only those exceptions to Regulatory Guide *l.97 that
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have been identified by the licensee. The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.
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3.
EVALUATION The Itcensee provided a response to Item 6.2 of NRC Generic Lattar 82-33 on April 30, 1984. Additional information was submitted on May 7, 1985 and August 23, 1988.
The responses describe the licensse's parition on post accident monitoring instrumentation. This evaluation is based on these submittals.
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1 3.1 Adherence to Reaulatory Guide 1.97 The licensee provided a review of their post-accident monitoring i
instrumentation that compareJ the instrumentation characteristics against the recommendations of Regulatory Guide 1.97, Revision 3.
The submittals provide an evaluation of Regulatory Guide 1.97 requirements and
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implementation plans for Browns Ferry. Therefore, we conclude that the licensee has provided an explicit comitment on conformance to Regulatory Guide 1.97, except for those deviations that were justified by the licensee as noted in Section 3.3.
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3.2 Tvoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific, manually controlled safety actions.
The licensee classifies the following instrumentation as Type A.
1.
containment hydrogen concentration 2.
drywell pressure 3.
drywell atmosphere temperature The instrumentation listed above meets the Category 1 recommendations consistent with the requirements for Type A variables.
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3.3 Excentions to Reaulatory Guide 1.97 The licensee identified the deviations and exceptions to Regulatory Guide 1.97 that are discussed in the following paragraphs.
3.3.1 Desian Cateoory Excentions Regulatory Guide 1.97 recommends that the following variables be monitored by Category 2 instrumentation. The licensee, in Reference 4 identified this instrumentation as Category 3.
The licensee, in Reference 5, clarifies the characteristics of the instrumentation provided and describes deviations from the recommended Category 2 environmental qualification.
Reference 6 provided additional clarification and commitments.
a.
Effluent radioactivity noble gases -- The Category 3 instrumentation was installed to and meets the requirements of NUREG 0737, Items !!.F.1-1 and II.F.1 2.
It is not exposed to harsh environments.
We find this to be a good faith attempt (as defined in NUREG-0737, Supplement No. 1. Section 3.7 (Reference 3)) to meet NRC requirements and is, therefore.
acceptable, b.
Primary system safety relief valve position -- The licensee's acoustic monitors for this' variable have been verified, in Reference 5, to be Category 2 instrumentation, c.
Reactor core isolation cooling (RCIC) flow -- The licensee states that the RCIC system is not an emergency core cooling system (ECCS), and that no credit is taken for it in the safety analysis of the plant.
Based on this statement, we find that Category 3 instrumentation is acceptable for this variable, s
d.
High pressyre coolant injection flow -- The licensee establishes, in Reference 5, that this variable is monitored by Category 2 instrumentation.
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Carutsprey flow.. The licensee comitted, in Referenco 6, to O.
provide environmentally qualified instruments for the FI-75-21 and
.I F1 75 49 instrument loops.
Based on this comitment, and the rest of instrument description describing Category 2 instrumentation, we find that the licensee's instrumentation and proposed upgrades satisfactorily meet the Regulatory Guide 1.97 recommendations.
f.
Low pressure coolant injection flow -- The licensea comitted, in Reference 6, to provide environmentally qualified instruments for the FI 74-50, FI 74-64, and FR 74 64 instrument loops.
Based on this commitment, and the rest of the instrument description describing Category 2 instrumentation, we find that the licensee's instrumentation and proposed upgrades satisfactorily meet the Regulatory Guide 1.97 recomendations.
g.
Standby liquid control system storage tank level -- The licensee verifies the accuracy of this instrumentation monthly using a gage stick, which can also be used post accident.
Procedures require
.the tank to be emptied once system operation is initiated.
The licensco states that no operator action is based on this
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indication.
Based on the described tank use, procedures, and alternate level determination, we find the Category 3 J
instrumentation acceptable, h.
Residual heat removal (RHR) system flow -- In Reference 6, the licensee described the RHR system and the subsystems for low pressure injection, drywell spray, and suppression pool spray.
The flow to any of these subsystems, and the RHR system flow is-measured by the same instrumentation.
The licensee comitted, in Reference 6, to provide environmentally qualified instruments for the FI 74 50, F1-74-64, and FR-74-64 instrument loops.
Based on this comitment, and the rest of the instrument description
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describing Category 2 instrumentation, we find that the licensee's instrumentation and proposed upgrades satisfactorily meet the Regulatory Guide 1.97 recomendations.
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1.
RHR heat exchaitger outlet temperature -- Regulatory Guida 1.97 recommends Category 2 instrumentation for this variable.
The licensee has provided instrumentation that, except for environmental qualification, is Category 2.
The licensee states (Reference 6) that, besides the heat exchanger outlet temperature, the torus (suppression pool) temperature, when trended, can be used to monitor the performance of the RHR heat exchangers.
Emergency operating instructions use the suppression pool temperature, suppression pool level, drywell temperature, drywell pressure, reactor level, and reactor coolant system pressure as the primary indicators for determining the heat energy remaining inside the primary containment.
This instrumentation, used for heat energy calculations, is Category 2 instrumentation.
Based on the above alternate instrumentation and associated emergency operating instructions, we find the described e
instrumentation for the RHR heat exchanger outlet temperature acceptable, j.
Cooling water temperature (to engineered safety features (ESF) system components -- Regulatory Guide 1.97 recommends Category 2 instrumentation with a range from 40'F to 200*F for this variable. The licensee states (Reference 6) that the Tennessee River is the source of cooling water for all ESF components.
The licensee describes Category 2 instrumentation, except for a cable that passes through the Unit 3 reactor building.
The licensee is committed to provide environmental qualification for this cable, bringing the entire instrument loop up to Category 2 requirements.
The indicator is located in the Unit 2 control room; there are no additional readouts in Unit 1 or Unit 3.
The licensee states that the river water temperature can only change slowly over a period of time, and because of this, the operators of any unit have timely access to the temperature information.
Regulatory Guide 1.97 makes allowance for displays outside of a 7
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eartts control room. We find this instance a suitable 'deviat' ion from the control room display requirements.
The range of the instrumentation is 90*F to 100'F.
The range recommended by the regulatory guide is much wider than this to allow for those units whose cooling water temperature can exceed the temperature of river water.
The licensee states that temperatures below 90'F are not of concern as the ESF i
components have a maximum design cooling water temperature of I
95'F.
The licensee also states that the river temperature and i
the cooling tower discharge channel do not exceed 95'F during plant operation.
Since there are no heat sources between the intake and the ESF components, there will be no significant change in water temperature. Cooling water flow can also be used to l
monitor system operation. We conclude that the instrument range is suitable for use with this variable.
k.
Cooling water flow to ESF system components --
Residual heat removal service water (RHRSW) flow -- The licensee committed, in Reference 6, to provide environmentally qualified instruments for the FI-23-36, FI-23 42, FI-23-48, and FI-23 54 instrument loops.
Based on this comitment, and the rest of the instrument description describing Category 2 instrumentation, we find that the licensee's instrumentation and proposed upgrades
-satisfactorily meet the Regulatory Guide 1.97 recommendations for the RHRSW flow instrumentation.
Emergency equipment cooling water (EECW) flow -- The licensee has not environmentally qualified the flow instrumentation for this variable and has requested a deviation from that recommendation for the EECW flow instrumentation. The EECW system is a system that is shared between the three units. There are four pumps, each powered by diesel generator-backed panelboards.
Each pump has circuit breaker status lights, pump motor ampere indication, and flow indication available in each control room. The circuit breaker status lights and pump motor ampere indication 8
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- instrumentation are located in mild environments, and are appropriate as alternate instrumentation for this variable.
The licensee states (Reference 6) that only two pumps are needed for
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full capacity of the EECW system.
The licensee also states that either of the two redundant systems can provide the complete cooling of the connected ESF heat loads because each component is supplied cooling water from both (via check valves) EECW headers.
The licensee also states that the EECW system valves are i
administrative 1y controlled in a preset position that apportions I
cooling water to the ESF components.
Based on the licensee's description of the EECW system and associated instrumentation, we find the deviation from the environmental qualification recommendations of Regulatory Guide 1.g7 for the EECW flow instrumentation acceptable.
1.
Emergency ventilation damper position - The licensee (Reference 6) revised the list of emergency ventilation dampers for each unit that have a safety function of controlling the radiological consequences of an accident. Ten of the dampers listed are in a mild environment, therefore, environmental qualification is not necessary; environmentally qualified position indication will be provided for twenty dampers; six dampers are being removed; and, in four cases, the damper is kept closed administrative 1y.
Based on the licensee's description, proposed modifications, and justification, we find the proposed instrumentation acceptable for use with this variable.
m.
Status of standby power -- The licensee identifies, in Reference 5, the indications used to control the diesel generator after a design basis accident and establishes that these indications are provided by Category 2 instrumentation, f'
n.
Noble gases and vent flow rate-comon plant vent -- This Category 3 instrumentation was installed to and meets the requirements of NUREG-0737, Items II.F.1-1 and II.F.1 2.
It is not exposed to harsh environments. We find this to be a good 9
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fhith attempt (as defined in NUREG 0737, Supplement. No.1, h.
Section 3.7 (Reference 3)), to meet NRC requirements and is, therefore, acceptable.
3.3.2 Neutron Flur Regulatory Guide 1.97 recomends Category 1 instrumentation for this variable.
The licensee has provided instrumentation that is not Category 1.
The licensee states that most portions of the neutron monitoring system are designed, procured and tested to standards more stringent than Category 3.
They identify the following components as not meeting Category I requirements:
source and intermediate range monitors (SRM and IRM) drive mechanisms and controls and the neutron monitoring system power sources.
The licensee concludes that, due to historical reliability and the redundancy of the overlapping channels (four SRM channels, eight IRM channels, and six average power range monitors), the currently installed instrumentation meets the intent of the guide. The power sources are the reactor protection system power supplies which are equivalent to Class 1E power.
l In Reference 6, the licensee states that a plant specific evaluation is pending the NRC acceptance of the Boiling Water Reactor Owners Group resolution to this concern.
During our review of the neutron flux instrumentation for boiling water reactors, we noted that the mechanical drives of the detectors and their cables generally have not satisfied the environmental qualification requirement of the Regulatory Guide 1.97. A Category 1 system that meets all the criteria of Regulatory Guide 1.97 has been an industry development item.
Based on our review, we conclude that the existing instrumentation is acceptable for interim operation. The NRC is currently reviewing the Owners Group position statement. The licasee should comit to comply with the results of this generic review.
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3.3'3 Coolant Level in Reactor Regulatory Guide 1.g7 recommends Category 1 instrumentation for this variable with a range extending from the bottom of the core support plate to the centerline of the main steamline or the top of the vessel (whichever is less). The licensee has instrumentstion that covers, with overlapping ranges, from 1/3 of the core height to above the centerline of the main steamline. Redundancy is provided up to 70 inches below the centerline of the main steamline. The shutdown vessel flooding range, which measures above this height, has no, redundancy and is Category 3 instrumentation.
The shutdown vessel flooding range instrument reference leg uses the top head vent as a penetration. In order to comply with the single failure requirement of Regulatory Guide 1.97, an additional head penetration would be needed for a redundant reference column for a second shutdown vessel flooding range channel.
Only the upper 70 inches of the recommended range is not monitored by Category 1 instrumentation.
The licensee notes that no manual or automatic functions are initiated in the upper 70 inches since these functions occur in the range monitored by Category I channels.
The licensee concludes that the reactor coolant level instrumentation meets the intent of the regulatory guide, and that only a marginal improvement in plant safety would be achieved by installing Category I shutdown vessel flooding range channels.
We find that an a second shutdown vessel flooding channel would not result
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in a significant increase in plant safety. We conclude that the single Category 3 shutdown vessel flooding channel is acceptable.
The licensee states that all automatic actions occur with the water level above the top of the active fuel.
The licensee also states that all manual actions to maintain adequate core cooling are the same anywhere below 2/3 of the core height.
Therefore, the licensee concludes that monitoring t
below 1/3 of the core height will not enhance the operator's action.
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&ased on the licensee's justification, we find this deviatien y - 2) ale, The licenses states that no recorder from a qualified instrument channel will be provided. This information is not essential for the operator's direct and imediate trend or transient information. However, a level recorder from a nonqualified instrument loop is provided for both ranges.
Item 6 of Table 1 of Revision 3 of Regulatory Guide 1.97 states that where direct and immediate trend or transient information is essential for operator inforwation or action, the recording should be continuously available on redundant dedicated recorders. Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.
Based on the licensee's statement that this information is not essential for the operator, we find that dedicated recorders are not i
required for this variable.
i 3.3.4 Reactor Coolant System Pressure The licensee states that they will not provide a recorder for the Category 1 reactor coolant system pressure channels.
Reactor pressure will be included in the database for the Safety Parameter Display System.
The justification provided by the licensee states that this information is not essential for the operator's direct and immediate trend or transient information. Howeve;, there is a pressure recorder in the control room that is part of a nonqualified instrument loop that the operators use during normal operation.
Based on the licensee's statement that the information is not essential for the operator, we find that dedicated recorders are not required for this variable.
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3.3.5 Drvww11 Snam level j
Drvwell Drains Sumo Level l
Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The sumps at Browns Ferry use level switches to initiate sump pump out. Timers indicate the duration of sump pump operation for estimating the amounts of leakage.
No safety related system is actuated either automatically or manually as a result of the sump level. The drywell sump systems are automatically isolated at the primary containment penetration should an t.ccident signal occur.
We conclude that the alternate instrumentation supplied by the licensee will provide appropriate monitoring for the parameters of concern.
This conclusion is based on the following.
a.
For small leaks, the alternate instrumentation is not expected to
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experience harsh environments during operation, b.
For larger leaks, the su:nps fill promptly and the sump drain lines isolate due to the increase in drywell pressure; thus negating the drywell sump level and drywell drain sumps level instrumentation, c.
This instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in a post accident situation.
Therefore, we find that the alternate Category 3 instrumentation provided is acceptable.
3.3.6 Primary Containment Pressure The licensee utilizes the instrumentation for the variable drywell-pressure to measure this variable.
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- iw The licensee states that the crywell and torus pressure may'not b'e a'
f er, mal at. an times; however, the drywell vacuum breakers have been designed to give assurance that torus pressure does not exceed drywell pressure by more than 2 psi. Also, the drywell pressure cannot exceed the torus pressure by greater than 1.5 psi due to the vent piping connections below the torus. water level. Drywell/ torus differential pressure is bounded by
-1.5 psi < AP < 2 psi. Therefore, drywell pressure can be used as an indication of torus pressure and drywell pressure provides the key variable for monitoring primary containment pressure.
Based on the justification provided by the licensee, we conclude that the alternate instrumentation supplied for this variable is adequate to monitor this variable during all accident and post-accident conditions.
l 3.3.7 Py,iggy Jytainment Isolation Valve Positiqn Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee has listed those valves that comply with this position and those that do not. Those that are identified as not complying are grouped below, with the justification following.
There are three valves (HCV-2-1383, HCV-33-1070, and HCV-74-722) that are locked c.losed.
The licensee indicates that surveillance is provided to assure that these valves are closed.
Therefore, we find this deviation acceptable.
There are two valve pairs (SCV-43-28A&B and SCV-43-29A&B) in the
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sampling and water quality system that are normally closed.
The licensee states that a local push button energizes two valves (A and B) simultaneously.
They fail closed, and when the pushbutton is released the valves close.
Flow through the sample line is observable at this sampling station. There it no control room control of these valves.
The licensee indicates that surveillance is provided to assure that these valves are closed. Therefore, we find this deviation acceptable.
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.c The isolation valve on the high pressure coolant injection line is
!f locked open. The licensee states that isolation is provided by a check valve. We find this deviation acceptable, as isolation is assured by the check valve.
There are 20 containment inerting system valves that are part of the hydrogen and oxygen monitoring system.
These valves do not have individual valve position indication. The licensee states that these valves are part of a qualified closed loop system that is designed to function under accident conditions.
The system is part of the containment boundary; the valves are not. The licensee indicates that surveillance is provided to assure that these valves are operable.
Each of two switches control the position of half of these valves; the valves are grouped in two redundant systems.
Each system has its isolation signal indicated; the isolation signal can be manually overridden by a manual key-locked switch.
Because these valves do not form a portion of the containment boundary, we find that lack of position indication.for these individual valves is acceptable, l
3.3.8 Bad.iAtlon Level in Circulatina Primary Coolant The licensee states that the critical actions taken to prevent and mitigate a gross breach of fuel cladding are to shut down the reactor and to maintain.the water level.
Neither of these are influenced by the recommended variable.
The applicant. indicates that the post-accident sampling facility (PASF) provides a means of obtaining samples of reactor I
coolant and determining the status of fuel cladding and that radiation monitors in the condenser off-gas and main steamlines provide information on the status of fuel cladding when the plant is not isolated.
Based on the justification and the' alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable. <
15
~
f 1.1.9 +PH mary containment Area Radiation--Hioh Rance 4
"' _ Regulatory Guide 1.97 recomends Category 1 instrumentation for this variable. The licensee is supplying Category 1 instrumentation.
The drywell electrical penetrations that are part of the instrument loops are not environmentally qualified.
These penetrations are to be re'placed with environmentally qualified electrical penetrations as scheduled in the Browns Ferry Integrated Schedule. Therefore, the instrumentation to be supplied for this variable is acceptable.
3.3.10 containment and Drvwell 0xvoen concentration Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable.
The licensee has Category 3 instrumentation. The licensee states that the function of detection of a potential for a breach in the L
_ containment is also monitored by the drywell pressure (Type A, Category 1),
- the drywell and torus hydrogen concentration (Type A, Category 1), and the reactor pressure (Type B, Category 1). The torus and drywell oxygen concentration is not used to initiate a safety function or to key the operator to perform a manual action.
Browns Ferry's primary containment is operated with an oxygen deficient (i.e., inerted) atmosphere as one part of those measures for combustible gas control.
The containment atmospheric dilution (CAD) system is used following a postulated loss-of-coolant accident (LOCA) to dilute the containment atmosphere with nitrogen to maintain the hydrogen and oxygen concentrations below combustible levels.
Hydrogen concentration is used at Browns Ferry to alert the operators to manually initiate the CAD System (hydrogen concentration is a Type A variable).
The oxygen concentration is used only as a surveillance instrument.
The licensee states that the NRC approved this configuration l
for Unit Nos. 1, 2, and 3 in technical specification amendments 38, 36, L
and 12 respectively.
i Based on the justification and the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied by the licensee for this variable is adequate and, therefore, acceptable.
16 L
r
p 3.3.'11 Co'ntainment Effluent Radioactivity i
Effluent Radioactivity Regulatory Guide 1.97 recommends the following instrumentation for noble gas:
for containment effluent radioactivity, Category 3 instrumentation with a range from 10-6 pCi/cc to 10*2 pCi/cc, for effluent radioactivity, Category 2 instrumentation with a range from 10-6 pCi/cc to 10 pCi/cc. The licensee has committed to 3
installing a system to monitor the Browns Ferry stack for high-range noble gas with particulate and iodine collection on appropriate collection media in response to NUREG-0737, Item II.F.1.1 and II.F.1.2.
This is stated in Reference 4.
The Browns Ferry plant is designed to have one designated release point; namely, the stack. The secondary containment features of the plant will isolate and/or realign to clearup systems. The cleanup systems exhaust to the designated release point. Therefore, there is a very low probability of a major release of activity within other plant zones such as the turbine building.
If an accidental release does occur in other areas, a high-radiation alarm is received and the affluent vent dampers and fans can be quickly isolated.
Since release paths such as the turbine building vents do not have cleanup systems, the licensee stetes that these ventilation system exhausts are isolated or shutdown, therefore, it is not necessary to determine quantitative releases from these points.
Reference 5 indicates that this instrumentation-is located in a non-harsh post-accident environment. Therefore, qualification to 10 CFR 50.49 is not required.
We therefore find the instrumentation 1
supplied for these variables acceptable.
3.3.12 Radiation Exoosure Rate Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 10*I R/hr to 104 R/hr. The instrumentation supplied by the licensee for this variable has a range of 10-1 mR/hr to 103 mR/hr 6
(10 mR/hr to 10 mR/hr in the transverse incore probe (TIP) room).
17
3-J*
[
- The licenseeljustifies the range by stating that, in general, access is nat. regn. tend te any area of the secondary containment in order to service
~,
equipment importent to safety _in a post-accident situation.
If and when accessibility is reestablished in the long term, it will be done by a combination of portable radiation survey instruments and post-accident sampling of the secondary containment atmosphere. This instrumentat h a has the recommended range, and will support long term surveillance and release assessment. The existing lower range area radiation monitors would be used only in those instances in which radiation levels were very mild.
From a radiological standpoint, if the radiation levels reach or exceed the upper limit of the ranges, personnel would not be pemitted into the areas without portable monitoring except for live saying.
Based on the alternate instrumentation used by the licensee for this variable, we find the proposed ranges for the radiation exposure rate monitors acceptable.
3.3.13 Sunoression Chamber Sorav Flow Drywell Sorav Flow Regulatory Guide 1.97 recommends Category 2 instrumentation with a range from zero to 110 percent of design flow for these variables. The licensee does not provide a direct measure for these variables.
The licensee states that the drywell sprays can be used to control the pressure and temperature of the drywell.
Likewise, the suppression pool sprays can be used to control the pressure and temperature in the torus. The flow to the sprays is monitored by a flow element which is common to both the drywell spray flow and the suppression pool spray flow.
This flow element is also used with the variables low pressure coolant injection (LPCI) flow-and RHR system flow.
The operator can determine that the indicated flow is the flow that is being diverted to the sprays by observing the position (in the main control room) of the valves in the residual heat removal (RHR) line. The effectiveness of these flows can be verified by pressure and temperature changes of the drywell and the torus. The drywell pressure and drywell temperature instrumentation have been classified as Category 1.
The torus pressure is nominally within 2 psi of the drywell pressure.
18
- t-4
' The instrumentation for these variables is being upgraded to Category 2 (Reference 6), satisfying the recommendations of Regulatory Guide 1.97.
3.3.14 Drywell Atmosohere Temeerature Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 40*F to 440'F.
The licensee has provided instrumentation for this variable with a range of zero to 400*F. They did not supply justification for not monitoring from 400'F to 440*F.
Examination of the Final Safety Analysis Report (FSAR), Figure 14.6-11, shows that the maximum post-accident drywell temperature is less than 300*F. Therefore, the range of zero to 400'F is adequate, and this deviation is acceptable.
3.3.15 Primary System Safety Relief Valve Position Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with indication of either closed-not closed, flow through the valve, or pressure (zero to 50 psig) in the valve discharge lines.
The licensee monitors this variable with two diverse methods, as indicated in Reference 4.
First, there is flow indication by acoustic monitors.
- Second, there is temperature indication of zero to 600*F for the valve discharge lines.
Reference 5 indicates that the acoustic monitors are Category 2.
Therefore, the instrumentation provided for this variable is acceptable.
3.3.16 Standbv Licuid Control System (SLCS) Flow The licensee has not provided flow instrumentation for this variable.
The licensee states that the purpose of the variable SLCS flow, as discussed in Regulatory Guide 1.97, is to monitor the operation of the SLCS.
SLCS operation is monitored in the control room with Category 3 tank level instrumentation, and by pump operation. When the squib valves are opened and SLCS pumps started, the total contents of the tank are pumped into the 19
i "reerter,1 FThu through the line to the reactor is indicated by in '
?E annunci'ntor and a white indicator light. The actual amount of flow to the L'
nauter is said to not be relevant because the total tank contents are to be pumped.
SLCS operation is monitored by tank level (decreasing), pump operation indication, and neutron flux response.
We find that the alternate instrumentation provided by the licensee will adequately monitor the variable of concern during all accident and post-accident conditions and is therefore acceptable.
3.3.17 Hiah Radioactivity Liouid Tank Level The licensee has not provided instrumentation for this variable.
The licensee states that the radioactive waste systems are designed to dispose of the radioactive process wastes generated during plant operation.
The system is designed to prevent the inadvertent release of significant quantities of radioactive material from the restricted area of the plant so-that resulting exposures are within the guideline values of 10 CFR 20.
The radwaste facility is located in the radwaste building.
The radwaste building has been designed to withstand a design basis earthquake (DBE).
Should the floor drain collector tank fail or overfill before isolation, the licensee states that any spilled liquid would be contained within the radwaste building.
The licensee's FSAR indicates that the lines discharging to the floor drain collector tank of the radwaste system from inside containment are automatically isolated on an accident signal. The spilled liquid would be retained in the building.
Because the leaks or spills from the radwaste system are retained within the radwaste building and have little or no effect on the site boundary dose rate and the radwaste system is not required after a design-basis accident (DBA), the level of the floor drain collector tank is not required to monitor the operation of the system.
Based on the licensee's justification, we find that monitoring this variable in the Browns Ferry control room is not necessary, 20
.-,,y
r Reactor Buildina or Secondary Containment Area Radiation s
3.3.18 The licensee states that this variable need not be implemented.
The licensee reports that the use of local radiation exposure rate monitors to.
detect breach or leakage through primary containment penetrations results in H
ambiguous indications.
This is due to the radioactivity in the primary containment, the radioactivity in the fluids flowing in emergency core coolant system piping and the amount and location of fluid and electrical penetrations. The licensee concludes that the use of the vent stack noble gas effluent monitors is the proper way to accomplish the detection of releases, release assessment, and long term surveillance reconmended for this variable. The licensee states that the vent stack noble gas effluent i
monitors cover the range recomended for this variable.
We find the l
alternate instrumentation provided is acceptable for this variable.
l l
3.3.19 Noble Gas and Vent Flow Rate--Common Plant Vent Regulatory Guide 1.97 recommends instrumentation for this variable.
The licensee states that they are committed to install a system to monitor the noble gas (high range) for the plant stack, which is the common plant release point. The licensee has identified the range of the instrumentation, and stated that it satisfies the requirements of NUREG 0737, Items II.F.1.1 and II.F.1.2.
t-Reference 5 indicates that this instrumentation is located in a non-harsh post-accident environment.
Therefore, qualification to 10 CFR 50.49 is not-required.
We therefore find the instrumentation supplied for this variable acceptable.
L 3.3.20 Particulates and Haloaens 1
Regulatory Guide 1.97 recommends instrumentation for this variable with f
a range of 10-3 pCi/cc to 102 #Ci/cc. The licensee is installing instrumentation.
They state that this instrumentation satisfies the requirements of NUREG-0737, Items II.F.1.1 and II.F.1.2.
In 21
l s*
4 lieference 5,. ttr isTs,hown to meet the recommendations of Regulatc.y "I
Guide 1.ST. We-find this instrumentation acceptable.
3.3.21 Airborne Radiohalocens and Particulates Rogulatory Guide 1.97 recommends portable sampling with onsite analysis 1
for this variable with a range of 10~9 pC1/cc to 10'3 pCi/cc.
The licensee is meeting this recomendation by laboratory analysis in accordance with NUREG-0737, Iten 11.8.3.
Reference 5 shows that this instrumentation meets the recommendations of Regulatory Guide 1.97. We find this instrumentation acceptable.
j 3.3.22 Plant and Environs Radiation Regulatory Guide 1.97 recommends portable instrumentation for this variable, with ranges of 10*3 R/hr to 10 R/hr photons and 4
10'3 rads /hr to 10 rads /hr beta radiation and low-energy photons.
The 4
licensee is supplying portable instrumentation for this variable.
Reference 5 states that the instrument range conforms with the recommendations of Regulatory Guide 1.97.
We find this instrumentation acceptable.
3.3.23 Accident Samolina (primary Coolant. Containment Air and Sumo)
-The licensee's. post-accident sampling facility provides sampling and analysis. However, there are deviations from the following recommendations.
1.
The sumps are not sampled.
2.
The licensee's submittal does not show compliance with the range recommendations.
f 3.
The containment air is not analyzed for hydrogen or oxygen content (continuous online monitors are used instead).
22
s n
' The'11censee takes exception-to Regulatory Guide 1.97 with respect to:
poit-accident sampling capability. This exception goes beyond the scope of this review and has been addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.
o 23
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9,. l,'.
4.
CONCLUSIONS a
'A:. % e en.
v Based on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 197, with the exception of the
- v3riable neutron flux. The licensee's instrumentation for neutron flux is acceptable on an interim basis until Category 1 instrumentation is developed and installed.
(Section3.3.2) r r
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5.
REFERENCES o
1.
Letter, NRC (D. G. Eisenhut) to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits,
" Supplement No. I to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2.-
Instramantation for Licht-Watea-Cooled Nuclone Power P' ants to Assess Plane. and Environs conditions )urina and :oowina an nec1ggni, Regu'atory Guide 1.97,-Revision 2, NRC, Office of Standards Development, December 1980.
3.
Clarification of TMI Action Plan Raouirements. Raouirements for Emeroonev Resoonse Caoability, NUREG-0737 Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4.
Letter, Tennessee Valley Authority (L. M. Mills) to NRC (H. R. Denton),
April 30, 1984.
5.
Letter, Tennessee Valley Authority (J. A. Domer) to NRC, May 7,1985.
6.
Letter, Tennessee Valley Authority (R. Gridley) to NRC, " Response to NRC's Safety Evaluation Report on Regulatory Guide 1.97 Compliance as Applied to Emergency Response Facilities Dated June 23, 1988,"
August 23, 1988.
7.
Instrumentation for Licht-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Fo110wina an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
25
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, com Division of Engineering and Systems Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 to. 5UPPLEMENTARY NOTts II. AS$TR ACT IJap eee w ous This EG&G Idaho, Inc., report documents the review of the Regulatory Guide 1.97, Revision 3 submittals for the Browns Ferry Nuclear Plant, Unit Nos. 1, 2, and 3, and identifies areas of nonconformance to the regulatory guide.
Exceptions to l
Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.
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