ML18092B583
| ML18092B583 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/30/1984 |
| From: | Felton J NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Garde B GOVERNMENT ACCOUNTABILITY PROJECT |
| Shared Package | |
| ML18092B584 | List: |
| References | |
| FOIA-84-779 NUDOCS 8502280520 | |
| Download: ML18092B583 (16) | |
Text
CONFORMANCE TO REGULATORY GUIDE 1.97 SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2
- l.
A. C. Udy Published January 1985 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-AC07-76ID01570 FIN No. A6483 Enclosure
!~,..*, '
ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, and identifies areas of full conformance to Regulatory Guide 1.97, Revision 2.
Any exception~ to these guidelines are evaluated and those areas where sufficient basis for. acceptability is not provided are identified.
FOREWORD This report is supplied as part of the "Program for Evaluating Licensee/Applicant Conformance to RG 1.97, 11 being conducted for the U.S.
~uclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Systems Integration, by EG&G Idaho, Inc., NRC Licensing Support Section.
The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11-3.
Docket Nos. 50-272 and 50-311 TAC Nos. 51128 and 51129 ii
CONTENTS ABSTRACT FOREWORD..............................................................
ii
- 1.
INTRODUCTION.....................................................
1
- 2.
REVIEW REQU I REM EN TS..............................................
2
- 3.
EVALUATION...........................................,............
4 3.1 Adherence to Regulatory Guide 1.97.........................
4 3.2 *Type A Variables................. *..........................
4 3.3 Except-ions to Regulatory Guide 1.97...* ~....................
5
- 4.
CONCLUSIONS......................................................
12
- 5.
REFERENCES*.......................................................
13
CONFORMANCE TO REGULATORY GUIDE 1.97 SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2
- 1.
INTRODUCTION On December 17, 1982, Generic Letter 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 caoabi 1 ity., These requirements have been published as Supprement 1 to NUREG-0737, 11 TMI Action Plan Requirements" (Reference 3).
The Public Service Electric and Gas Company, the licensee for the Salem Nuclear Generating Station, provided a response to the generic letter on April 15, 1983 (Reference 4).
The letter referred to a previous letter dated April 2, 1981 (Reference 5) for a review of the instrumentation provided for Regulatory Guide 1.97.
The licensee provided additional information for this review in letters dated September 21, 1983 (Reference 6) and August 9, 1984 (Reference 7).
This report provides an evaluation of these submittals.
l
- 2.
REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee meet~
the guidance of Regulatory Guide 1.97 as applied to emergency response facilities.
The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97:
- 1.
Instrument range
- 2.
Environmertal qualification
- 3.
Seismic qualification
- 4.
Quality assurance
- 5.
Redundance and sensor location
- 6.
~6wer supply
- 7.
Location of display
- 8.
Schedule of installation or upgrade.
Further, the submittal should identify deviations from the guidance in the regulatory guide and provide supporting justification or alternatives.
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 matter.
At these meetings, it was noted that the NRC review would only address exceptions taken to the guidance of Regulatory Guide 1.97.
Further, where licensees or applicants explicitly state that instrument systems conform to 2
the provisions of the Guide it was noted that no further staff review would be necessary.
Therefore, this report only addresses exceptions to the guidance of Regulatory Guide 1.97.
The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.
3
- 3.
EVALUATION The licensee provided a response to the NRC Generic Letter 82-33 on April 15, 1983.
This response referred to an earlier submittal of April 2, 1981, which described the licensee's position on post-acci~ent monitoring instrumentation.
Additional information was provided on September 21, 1983, and August 9, 1984.
This evaluation is based on these submittals.
3.1 Adherence to Regulatory Guide 1.97 The licensee stated that the guidance of Regulatory Guide 1.97 has been-implemented.
Confor_mance includes instrumentadon that meets the guidance, and instrumentation that was added or modified to meet the g~idance.
Instrumentation that is not fully in compliance, but where the licensee views it as appropriate for the variable, and items which are not part of the station design were noted.
Therefore, it is concluded that the
- --lftensee has provided an explicit commitment on conformance to the guidance of Regulatory Guide 1.97, except for those exceptions that were justified as noted in Section 3.3.
3.2 Type A Variables In that Regu~atory Guide 1.97 does not specifically identify Type A variables, i.e., those variables tha~ pruvide information required for operator controlled safety actions, the licensee classified the following instrumentat~on channels as Type A variables:
- 1.
Reactor coolant system hot leg water temperature
- 2.
Reactor coo1ant system pressure
- 3.
Degrees of subcooling
- 4.
Containment pressure 4
- 5.
Effluent radioactivity--noble gas effluent from condenser air removal system exhaust
- 6.
Refueling water storage tank level
- 7.
Pressurizer level
- 8.
Steam generator pressure
- 9.
Auxiliary feedwater flow
- 10.
Condensate storage tank water level
- 11.
Steam generator blowdown radiation.
All of the above variables, except number 11, are also included as Type B, C or D variables.
All meet Category 1 requirements consistent with the requirements for Type A variables.
3.3 Exceptions to Regulatory Guide 1.97 The licensee identified the following exceptions to the requirements of Regulatory Guide 1.97.
3.3.1 Reactor Coolant System Cold Leg Water Temperature The licensee has provided instrumentation for this variable that satisfies the recommendations of Regulatory Guide 1.97 except that the range is 0 to 700°F rather than the 50 to 750°F recommended by Revision 2 of the regulatory guide.
The licensee indicates that the range supplied covers all accidents except where the reactor coolant becomes superheated.
Revision 3 of Regulatory Guide 1.97 (Reference 7) recommends a range of 50 to 700°F, which is met by the supplied instrumentation.
Therefore, there is no r:10vi<'\\tion from the c:urr~nt Pevision of th~ regulatory guide.
5
- e.
3.3.2 Reactor Coolant System Hot Leg Water Temperature*
The licensee has provided instrumentation for this variable that satisfies the recommendations of Regulatory Guide 1.97 except that the range is 0 to 700°F rather than the 50 to 750°F recommended by Revi-sion 2 of the regulatory guide.
The licensee indicates that the range supplied covers all accidents except where the reactor coolant becomes superheated.
Revision 3 of the regulatory guide recommends a range of 50 to 700°F, which is met by the supplied instrumentation.
Therefore, there is no deviation from t~e current r*evision of the regulatory guide.
3.3.3 Radiation Level in Circulating Primary Coolant Regulatory Guide 1.97 recommends instrumentation for this variable for the detection of a breach.
The licensee has provided radiation monitoring on the letdown line.
The letdown line is isolated for an accident situation.
The licensee would then utilize the post-accident samoiing system, which is available with the reactor isolated.
We concur with the justification submitted by the licensee for this deviation.
Their existing instrumentation is adequate to monitor post-accident reactor coolant activity.
Further, a continuous post-accident reactor coolant activity monitor is not a require~ent of NUREG-0737.
Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.A Residual Heat Removal Heat Exchanaer Outlet Temperature The licensee indicates that the instrumentation for this ~ariable has no seismic or environmental qualification test data available.
Our review of the requirements of Regulatory Guide 1.97 for Cat~gory 2 instrumentation shows that seismic qualification is not required.
The licensee states that the operator does not use this instrumentation during an accident.
Additionally, they indicate that the RCS cold leg wa~er temperature provides the same information.
6
Environmentai qualification has been subsequently clarified by the environmental qualification rule, 10 CFR 50.49.
It is concluded that the guidance of Regulatory Guide 1.97 has been superseded by a regulatory requirement.
Any exception to this rule is beyond the scope of this re~iew and should be addressed in accordance with 10 CFR 50.49.
3.3.5 Accumulator Tank Level and Pressure Regulatory Guide 1.97 recommends a range for this variable of 10 tc 90 percent of volume and 0 to 750 psig.
The licensee has identified a deviation in that the level instrumentation for this variable covers a range of 52.65 to 7Q.29 percent of volume.
The licensee's justification for this deviation is that the present range is needed to meet the instrument accuracies required by technical specifications to ensure an adequate volume of borated water before any less-of-coolant accident.
The accumulators are passive and automatically discharge for reactor coolant system (RCS) breaks.
The level and pressure measurement channels are not required to protect the integrity of the RCS boundary, to shut down the reactor.or to maintain it in a safe shutdown condition or to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
We find that the instrumentation supplied for this variable (level and pres5ure) is adequate to determine that the accumulator~ have discharged.
Therefore, this instrumentation is acceptable for this variable.
3.3.6 Refueling Water Storaae Tank Level The licensee has supplied instrumentation for this variable that covers a range of 2.5 to 45.24 ft of the 48 ft tank height.
The regulatory guide specifies a span of top to bottom.
The licensee indicates that the tank overflow is at 45.24 ft.
Therefore, the upper limit of the span is the effective tcp of the tank. *The licensee indicates that the top of the tank discharge line is 1.83 ft from the bottom of the tank.
The tank is P5~~ntially empty at 2.5 ft.
The difference between 1.83 and 2.5 ft is 7
1.5 percent of the tank height; this is within the accuracy of the instrumentation.
Therefore, we conclude that the instrumentation range supplied for the refueltng water storage tank level is acceptable.
3.3.7 Pressurizer Level The licensee has supplied instrumentation for this variable with a range of 4 ft 10 in. to 48 ft 6 in.
Regulatory Guide 1.97 specifies a range of top to bottom.
The hemispherical ends of the pressurizer (where the height to volume ratio is non-linear) are not measured.
The licensee provides the following justification for 'this deviation:
_a.
The range being monitored is 84.4% of the total height of the pressurizer.
- b.
It provides the required information for the operator to take the necessary corrective action during a transient.
- c.
The minimum water level indicated is 11 feet 3-1/8 inches which is above the electric heaters.
- d.
The range being monitored in terms of percentage of total pressurizer height is 9.3% to 93.7%.
We concur with the licensee that the range of the pressurizer level instrumentation is adequate.
3.3.8 Quench Tank Level The licensee has provided instrumentation for this variable with a range of 7 in. to 8 ft 11 in. out of a total height of 9 ft 6 in.
Regulatory G~ide 1.97 recommends that the full height be covered by the instrument range.
The licensee indicates that the range adequately covers from 5 to 95 percent of the tank volume.
We concur with the licensee that the range of the quench tank level instrumentation is adequate.
8
3.3.9 Quench Tank Temoerature The licensee has supplied instrumentation for this variable that has a range of 50 to 350°F instead of the recommended 50 to 750°F.
The licensee states that the tank rupture disk has a design pressure of 85 psig,* and that this restricts the temperature of the saturated steam to 328°F.
The pressure would have to reach 134 psig for the temperature to exceed the range of 350°F.
We concur with the licensee 1s analysis and find that this deviation is acceptable.
3.3.10 Steam Generator Level
- The l~censee has supplied instrumentat~on for* this variable that measures from 12 in. above the tube sheet to 587 in. above the tube sheet (this is in the separators).
Regulatory Guide 1.97 recommends instrumentation with a range from the tube sheet to the separators.
At 12 in. above the tube sheet (2 percent of the range), the steam generator is essentially empty.
We view this deviation in range as minor, and, therefore, acceptable.
3.3.11 Contain~ent Spray Flow Regulatory Guide 1.97 recommends instrumentation for this variable to monitor operation of the containment spray.
It recommends Category 2 instrumentation with a range from 0 to 110 percent of design flow.
The licensee has not provided a direct measurement of containment spray flow.
Instead they use an indirect measurement of the spray additive flow.
The licensee has stated that this instrumentation 11meets the (Category 2) requirements of Regulatory Guide l.97.
11 The additive flow is proportional to the containment spray flow except when the additive tank is depleted.
Then the pump motor current and discharge valve position will i~dicate system operation (but not the actual flow).
9
We concur with the licensee that this alternate instrumentation is adequate to monitor the operation of the containment spray system.
3.3.12 Containment Sump Water Temperature Regulatory Guide 1.97 recommends instrumentation for this variable to monitor operation of the containment cooling systems.
The licensee justifies not monitoring this variable in the sump by stating that 11 emergency core cooling and containment heat removal system pumps, specifically the residual heat removal pumps which take suction from the containment sump when the refueling w~ter storage tank is empty, were designed to meet the criteria in Safety Guide 1.
11
~afety Guide 1
- *(Reg-ulatory Guide-Ll), when followed, provides adequate net positive suction head to the pump.that.draw suction from the containment sump, assuming maximum expected temperature of the sump contents with normal
(~.e., minimum) ambient containment pressure.
The licensee monitors the residual heat removal heat exchanger inlet temperature.
This temperature is indicative of the sump water temperature once recirc~lation of the sump contents begins.
We find that this alternative is-acc.eptable.
3.3.13 Volume Control Tank Level Regulatory Guide 1.97 recommends instrumentation for this variable tc monitor operation of the chemical and volume control system.
The licensee provides instrumentation for this variable-that measures from 16.5 to 85 percent of total volume instead of the regulatory guide recommended top to bottom.
The tank overflow line is at a level equivalent to 85 percent of total volume.
Thus the upper limit of the range is at fu11 volume.
The licensee states that the range is adequate for the requirements of their technical specifications, and that this instrumentation is not required for an accident.
Section 9.2.3.1 of the FSAR (Reference 8) confirms this--the volume control tank is automatically valved off with an accident signal.
Based on this, we concur that the licensee's justification for this deviation i~ acceptable.
10
3.3.14 Component Cooling Water Flow to ESF System Regulatory Guide 1.97 recommends instrumentation for this variable to insure that the ESF equipment is supplied with adequate cooling water. ihe licensee indicates that this instrumentation satisfiis the specifi~ations of Regulatory Guide 1.97 except that it provides useful infor~ation only 11 during periods of recirculation." The component cooling water system is an intermediate system between the reactor coolant and the engineered safety feature (ESF) systems and the service water system.
It is ORerated in a closed loop mode.
It is manually aligned to the ESF equipment.
Therefore, we find that the licensee 1 s instrumentation is suitable for post-accident monit9ring of this variable.
3.3.15 Radioactive Gas Holdup Tank Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 0 to 150 percent of design pressure.
The licensee has instrumentation that reads from 0 to 150 psig rather than 150 percent of design pressure.
The operating pressure of the tanks and compressors is 110 psig.
The tank is isolated at this pressure automatically.
An alarm sounds should a tank pressure reach 135 psig.
Additionally, each tank has a pressure relief valve set at 150 psig.
We find that the range of the radioactive gas holdup tank pressure instrumentation is adequate.
11
- 4.
CONCLUSIONS Based on our review we find that the licensee either conforms to or is justified in deviating from the guidance of Regulatory Guide 1.97 with tne following exceptions:
- 1.
RHR heat exchanger outlet temperature--environmental q~alification should be applied in accordance with Section (g) to 10 CFR 50.49 (Section 3.3.4).
12
- 1.
- 2.
- 3.
- 4.
5,*
- 6.
- 7.
- 8.
- 5.
REFERENCES NRC letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction
- Permits, 11 Supp1 ement No. 1 to NUREG-0737--Requi rements for Emergency Response Capability (Generic Letter No. 82-33),
11 December 17, *1982.
Instrumentation for Lioht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Follcwing an Accident, Regulatory Guide 1.97, Revision 2, U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1980.
Clarification of TMI Action Plan Requirements, Requirements for Emergency Response Capability, NUREG-0737 Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
Public Service.Electric and Gas Company le:.ter,* E. A. Liden to
-Director of-tiluclear Reactor Regulation, NRC, Requirements for Emergency Response Capability, Supplement 1 to NUREG-0737, Preliminary Status Report and Schedule, 11 April 15, 1983.
Public Service Electric and Gas Company letter, R. L. Mittl to Director of Nuclear Reactor Regulation, NRC, 11 Compliance with Regulatory Guide 1.97, No. 2 Unit, 11 April 2, 1981.
Public Service Electric and Gas Company letter, E. A. Liden to Director of Nuclear Reactor Regulation, NRC, "Compliance with Regulatory Guide 1.97, NRC Request for Additional Information,u September 21, 1983.
Public Service Electric and Gas Company letter, E. A. Liden to Office of Nuclear Reactor Regulation, NRC, "Conformance to Regulatory Guide 1.97, Requirements for Emergency Response Capability, 11 August 9, 1984.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
- 9.
Salem Nuclear Generating Station, Units l and 2, Final Safety Analysis Report, Public Service Electric and Gas Company, Newark, NJ, August 27, 1971, Amendment 10.
37070 13