ML19312D715
| ML19312D715 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/21/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312D712 | List: |
| References | |
| NUDOCS 8003250235 | |
| Download: ML19312D715 (11) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 24 TO FACILITY OPERATING LICE _NSE NO. DPR-70 PUBLIC SERVICE ELECTRIC AND aAS COMPANY, PHILADELPHIA ELECTRIC COMPANY, DELMARVA POWER AND LIGHT COMPANY, AND ATLANTIC CITY ELECTRIC COMPANY SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NO. 50-272 Introduction By letter dated 0:tober 26,1977 (Reference 5) Public Service Electric and Gas Company (PSEG) submitted to the NRC a plant specific analysis in support
'of the proposed reactor vessel overpressure mitigating system (OMS) for Salem Nuclear Plant Unit 1.
This information supplements other documentation previously submitted by PSEG (References 2-4, 7).
The NRC staff has completed its review of all information submitted by PSEG in support of the proposed overpressure mitigating system and has found that the system provides adequate protection from overpressure transients.
A detailed safety evaluation follows.
PSEG, by Reference 8, has proposed for Salem Unit No.1 the applicable sections of the Standard Technical Specifications developed for the Salem Unit No. 2 OMS and approved by the staff.
Discussion Over the last few years, pressure transient incidents have occurred in pressurized water reactors. As used in this report, " pressure transient" is an event during which the Technical Specification temperature / pressure limits of the reactor vessel are exceeded. All of these incidents occurred at relatively low temperature (less than 200*F) where the reactor vessel material toughness (resistance to brittle failure) is reduced.
The " Technical Report on Reactor Vessel Pressure Transients" in NUREG-0138
-(Reference 6) surmarizes the technical considerations relevant to this matter, discusses the safety concerns and existing safety margins of operating reac'wrs, and describes the regulatory actions taken to resolve this issue by recucing 'the likelihood of future pressure transient incidents a t operating reactors.
A brief discussion is presented here.
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. A.
Vessel Characteristics Reactor vessels are constructed of high quality steel made to rigid specifications, and fabricated and inspected in accordance with the time-proven rules of the ASME Boiler and Pressure Vessel Code.
Steels used are particularly tough at reactor operating conditions.
However, if subjected to high pressures at low temperatures, these steels are less tough and could possibly fail in a brittle manrier.
Accordingly, power reactors have always operated with restrictions on the pressure during cold startup and cold shutdown operations when the steel temperatures are relatively ~ )w.
At operating tenperatures, the pressure allowed by Ap'pendix G limits is in excess of the relief setpoi.-t of currently installed pressurizer code safety valves. However most operating PWRs did not have pressure relief devices to prevent pressure transients that would exceed Appendix G limits during cold conditions.
B.
Regulatory Actions By letter dated August 27, 1976 (Reference 1) the NRC requested that PSEG begin efforts to design and install plant systems to mitigate the consequences of pressure transients at low tenperattres.
It was also requested that operating procedures be examined and administrative changes be made to guard against initiating pressure transients.
It was felt by the staff that proper administrative controls were required to assure safe operation for the period of time prior to installation of the proposed overpressure mitigating hardware.
PSEG. responded (Reference 2) with preliminary information describing interim measures to prevent these pressure transients along with some discussion of proposed hardware.
PSEG proposed to install hardware to provide a low pressure actuation setpoint on the existing pressurizer air operating relief valves.
PSEG participated as a member of a Westingbouse user's group which was formed to support the analysis effort required to verify the adequacy of the proposed system to prevent pressure transients.
Using input data generated by the user's group, Westinghouse per-formed trensient analyses (Reference 7) which are used as the basis for each plant specific analysis.
Die NRC staff requested additional information concerning the pr oposed procedural and hardware changes.
PSEG provided the required re ponses (References 3 and 4).
Reference 5 transmitted the plant specif c analysis for Salem Unit 1.
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. C.
Design Criteria Through this series of meetings and correspondence with PWR vendors and licensees, the NRC staff developed a set of criteria for an acceptable overpressure mitigating system.
The basic criterion is that the citigating system will prevent reactor vessel pressures in excess of these allowed by Appendix G.
Specific criteria for system performance are:
(1) _ Operator Action:
No credit can be taken for operator action for ten minutes after the operator is aware of a transient.
(2) Single Failure: The system must be designed to relieve the pressure transient given a single failure in addition to the failure that initiated the pressure transient.
(3) Testability: The system must be testable on a periodic basis consistent with the system's employment.
(4) Seismic and IEEE 279 Criteria:
Ideally, the system should meet seismic Category I and IEEE 279 criteria.
The basic objective is that the system should not be vulnerable to a common fa.ilure that would both initiate a pressure transient and disable the overpressure mitigating system.
Such events as loss of instrument air and loss of offsite power must be considered.
The HRC staff also instructed the licensee to provide an alarm which monitors the position of the pressurizer relief valve isolation valves, along with the low setpoint enabling switch, to assure that the overpressure mitigating system is properly aligned for shutdown conditions.
D.
Design Basis Evrqts The incidents that have occurred to date have been the result of operator errors or equipment failures.
Two varieties of pressure transients can be -identified: a mass input type from charging pumps, safety injection pumps, safety injection accumulators; and a heat adcition type which causes thermal expansion from sources such as steam generators or decay heat.
On Westinghouse design plants, the most common cause of the pressure transients has been improper isolation of the letdown path.
Letdown during low pressure operations is via a flowpath through the RHR, system.
Isolation of RHR with a charging pump running can initiate a pressure transient. Although other pressure transients occur with lower frequency, those _which result in the most. rapid pressure increases were identified by the NAC staff for analysis.
The most limiting nass input transient we identified is inadvertent injection by the largest safety injection pump.
The most limiting thermal
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s-expansion trar.sient is the start of a reactor coolant pump with a 50*F tsaperature difference between the water in the reactor vessel and the water in the steam generator.
Based on the historical record of pressure transients and the impositior. of more effective administrative controls, the NRC staff believes that the limiting events identified above form an acceptable bases for analysis of the prouosed overpressure mitigating system.
System Descriction and Evaluation PSEG's overpressure protection system (pressurizer overpressure protection system - POPS) is similar to the " Reference Mitigating System" developed by Westinghouse ar.d the user's group as described in Reference 2.
PSEG has proposed to nodify the actuation circuitry of the existing air operated pressurizer relief valves to provide a low pressure setpoint at 375 psig during startu; and shutdown conditions. When the reactor vessel is at low temperatures, with the low pressure setpoint selected, a pressure transient is terminated below the Appendix G limit by automatic opening of these relief valves. A manual switch is used to enable and disable the led setpoint of each relief valve. An enabling alarm which monitors reactor coolant system pressure, the position of the enabling switch and the upstream isolation valve is provided.
The POPS with its low presscre setpoint is enabled at a RCS temperature of 312*F during plant coolcown and is disabled at the same temperature during plant heatup. The NRC staff finds PSEG's POPS to be an acceptable concept for an overpressure mitigating system.
Discussion and evaluation of this system follows.
A.
Air Supoly The Salem clar.thas two power operated relief valves (PORVs). They are gate valves that are spring closed and air opened.
Each of the two PORVs receives actuating air from one of three available sources:
two fully redundant control air headers and a backup air accumuletor.
The accumulators are sized to provide sufficient actuating air for up to 100 cycles of PORV opening and closing (about ten minutes of operation; shculd the normal air supplies both become unavailable.
l In the event cf a Control air header rupture or other low pressur'e condition, ser. sors act to reposition valves to isolate the affected portion o' the system and automatically align a backup air supply.
Pressure alarr.s are installed in the control room to alert the plant operators tc a low air pressure condition.
The staff finds the PORV actuating air supply system for Salem Unit 1 acceptable.
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-S-B.
Electrical Controls A Technical Evaluation Report that was prepared for us by the Lawrence Livermore Laboratory as part of our technical assistance program is attached to this Safety Evaluation as Appendix A.
The. POPS design infonnation detailed in this section was derived from Reference 5.
The design for Salem Unit 1 POPS is a two-train system which uses separate and independent pressure transmitters to open the two pressurizer PORVs (lPRI and 1PR2) in the event that RCS press::re exceeds 375 psig.
This automatic action takes place provided the system has been manually enabled by placing two key -
locked pushbuttons in the "on" position.
Each PORV is actuated by its own logic relay which is energized by a bistable device. The bistable device is energized when the RCS pressure exceeds 375 psig.
The existing installed pressure sensors are used to develop the signal for valve actuation.
These are the same senscrs which provide the signal for automatic closure of the residual reat removal (RHR) suction paths at 600 psig. We find this design acceptable.
C.
Tes tability Testability will be provided.
PSEG has stated that verification of operability is possible prior to RCS low temperature operation by use of the recotely operated isolation valve, enable / disable switch and nomai electronics surveillance methodology.
Testing requirements have been proposed for incorporation in the Technical Specifications as discussed elsewhere in this evaluation.
Appendix G Cu ve The Appendix 5 curve submitted by PSEG for purposes of pressure transient analysis is based on 13 effective full power years irradiation.
The zero degree hsatup curve is allowed since most pressure transients i
occur during isethermal metal conditions. Margins of 60 psig and 10*F l
are included Sr possible instrument errors. The Appendix G limit at ll*F accordin; to this curve is 460 psig. The NRC staff finds that use i
of this curve is accetpable as a basis for oetermining proper POPS perfonnance.
Setooint Anal sis The one loop version of the LOFTRAN (Reference WCAP 79-07) code was used to perfo-n the mass input analysis. The four loop version was used for the heat input analysis.
Both versions require some input k
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O 4 rmdeling and initialization changes were required.
LOFTRAN is currently under review by the NRC staff and is judged to be an acceptable code for treating problems of this type.
The results of this analysis are provided in terms of PORV setpoint overshoot.
The predicted maximum transient pressure is simply the sum of the overshoot magnitude and the setpoint magnitude.
The PORV setpoint is adjusted so that given the setpoint overshoot, the resultant pressure is still below that allowed by Appendix G limits.
PSEG presented the following Salem Unit 1 plant characteristics to determine the pressurc reached for the design basis pressure transients:
SI Pump Flowrate 0 500 psig 108.4 lb/sec 3
RCS Volume 12,800 ft PORV Opening Time, Setpoint 2 sec, 375 psig 2
SG Heat Transfer Area 51,500 ft Westinghouse identified certain assumptions used in LOFTRAN that are conservative and tend to overpredict the peak RCS pressure in the design base transients. These are listed below along with some piant parameters Westinghouse has assumed in the generic analysis tr.at the licensees has identified to be conservative relative to the actual Salem Unit i values.
(1) One PORV was assumed to fail.
(2) The RCS was assumed to be rigid with respect to metal expansion.
(3) No credit was taken for the reduction in reactor coolant bulk modulus at RCS temperatures above 100*F (constant bulk modulus at all i25 temperatures).
(4) No credit was taken for the shrinkage effect caused by low temperature SI water added to higher temperature reactor coolant.
(5) The entire volume of water of the steam generator secondary was assumed available for heat transfer to the primary.
In reality, the liquid imediately adjacent and above the tube bundle would be the primary source of energy in the transient.
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.O 7-(6) The overall steam generator heat transfer coefficient, u, was assumed to be the free convective heat transfer coefficient of the secondary hsec-The forced convective heat transfer coefficient of the primary, h ri and the tube metal resistance p
has been ignored thus resulting in a conservative (high) coefficient.
(7) The RCP startup time assumed in the heat input analysis was 1,64 see whereas the actual SI pump startup time is 5.0 sec.
.The staff agrees th'at most of these assumptions are conservative.
It is prudent is assume the failure of one PORV.
A.
Mass Input Case The inadvertent start of a safety injection pump with the plant in a cold shutdown condition was selected as the limiting mass input Case.
Westinghouse provided PSEG with a series of curves based on the LOFTRAN analysis of a generic plant design which indicates PORV setpoint overshoot for -this transient as a function of system volume, relief valve opening time and relief valve setpoint.
These sensi-tivity analyses were then applied to the Salem Unit 1 plant parameters to obtain a conservative estimate of the PORV setpoint overshoot.
The staff finds this method of analysis to be acceptable.
Using the Westinghouse methodology, the SLlem Unit 1 PORV setpoint overshoot was determined to be 71 psi. With a relief valve setpoint of 375 psig a final pressure of 446 psig is reached for the worst case mass input transient.
Since the 13 EFPY Appendix G curve limit at temperatures above 100*F is 460 psig, we concluded that the system performan:e is acceptable with a 375 psig low pressure relief valve setpoint.
B.
Heat Inou: Case Inadverter.t startup of an RCP.with a reactor coolant to secondary coolant temperature differential across the steam generator of 50'F, and with the plant in a, water solid condition, was selected as the limiting heat input case.
For the heat input case Westinghouse provided ?SEG with a series of curves based on the LOFTRAN analysis of a generic plant design to determine the PORV setpoint overshoot -
as a-func icn of RCS volume, steam generator UA and initial RCS
. tempera ture.
For this transient, the reference PORV selected was assumed to have a totaltopening time of three seconds from the
~ instant signal.to open is received until the valve reached the full-open posi:'on.
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.The calculated maximum RCS pressure for the. heat input transient for a fixed AT of 50*F depends on the initial RCS temperature and is given here:
Initial RCS Temperature Maximum RCS Pressure 100*F 416.2 psig 180*F 442.2 psig 250*F 456.5 psig In none of these cases, is the Appendix G curve limit exceeded.
The NRC staff, therefore, finds that the analyses of both the limiting mass input case and the limiting heat input cases show maximum pressure transients below those allowed by Appendix G curve limit.
Imolementation Schedule The Salem Unit 1 POPS was installed and tested (pre-operational check-out) during the refueling outage ending in November 1977.
The system is fully operational.
Operatino procedures To supplement the hardware modifications and to limit the magnitude of postulated pressure transients to within the bounds of the analysis provided by PSEG a defense in depth approach is adopted using procedural and administrative ~ controls.
A number of provisions for prevention of pressure transients are contained in the Salem Unit 1 operating procedures. The procedures for startup (and jogging) a reactor coolant pump require that a steam bubble is established in the pressurizer prior to pump start, or the SG/RCS 4
AT be verified to be less than 50*F.
Also, shutdown procedures have been revised to include provisions for maintaining a steam-bubble in the pressurizer during plant cooldowns.
For conditions that do not require opening of the RCS a low pressure steam bubble will be maintained during the cold shutdown conditions.
By following these procedures,- PSEG does not anticipate that the RCS
.will be operated or maintained in a water solid condition except during RCS--fill and vent procedures.
-During plant cooldowns,- the power to both. safety injection pumps (Sips) is removed by racking out the power supply breakers when the RCS temp-erature is below 350*F. Also, SI header isolation valves are shut and thei r' power is - removed. The_ SIPS are de-energized whenever the RCS a
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9 temperature is below 312*F except when a special surveillance test is being conducted. During these procedures, only one SIP is energized; thus the POPS is able to keep RCS pressure below the Appendix G limit should an inadvertent mass addition from the single SIP occur during this procedure.
The staff finds that the procedural and administrative controls described are acceptable. However, the staff has detennined that certain procedural and administrative controls should be included in the Techni:al Specificatioas. These are listed in the following section.
Technical Spe:ifications To assure operation of the overpre.
-e mitigating system, the licensee has accepted for Salem Unit No.1 the staff's format for Standard
- Technical Spe:ifications and identical wording of the applicable sections of the Techni:al Specifications developed for Salem Unit No. 2 (Reference 8). -These changes to the. Technical Specifications are consistent with the intent of the statements listed below.
1.
3oth PORVs cust be operable whenever the RCS temperature is less than the r.ininum pressurization temperature (312*F), except one PORY ::ay be inoperable for seven days.
If these conditions are not met, the reactor coolant system must be depressurized and vented to the atmosphere or to the pressurizer relief tank within eight hou s..
2.
Operability of POPS requires that the low pressure setpoint be selected, the upstream isolation valves open and the backup air supply charged.
3.
No-mere t an one high. head SI pump may be energized at RCS temperatures below 312 F.
' 4 A reactor.ccolant pump may be started (or jogged) only if there is a stear. bubble in the pressurizer, or if the SG/RCS AT is less than 50*F.
- 5..The POPS ::usc be tested on a periodic basis consistent with the need for f ts use.
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Summary The administrative controls and hardware changes proposed by Public Service Elect-ic and Gas Company provide protection for Salem Unit 1 K
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. from pressure transients at low temperatures by reducing the probability of initiation of a transicrt and by limiting the pressure of such a transient to below the limits set by Appendix G.
The NRC staff finds that the overpressure mitigating system proposed by PSEG meets the
- criteria established by the NRC and is acceptable as a long term solution to the problem of pressure transients. Any future revisions of Appendix G limits for Salem Unit.1 must be considered and the overpressure mitigating system setpoint adjusted accordingly with corresponding adjustments
= in the license.
Environmental Consideration
. We have determined that the amendment does not authorize a change in effluent. types or total amounts nor an increase in power level and will not-result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have conciuded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amandment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be concucted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and cecurity or to the health and safety of the public.
Da te: February 21, 1980 S
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. References 1.
NRC (Kniel). letter to Public Service Electric and Gas Company, (PSEG) dated August 27, 1976.
2.
PSEG (Librizzi) letter to NRC (Kniel) dated October 25, 1976.
3.
PSEG (Librizzi) letter to NRC (Lear) dated March 25, 1977.
4.
PSEG (Librizzi) letter to NRC (Lear) dated May 3,1977.
- 5..PSEG (Librizzi) letter to NRC (Lear) dated October 26, 1977.
6.
" Staff Discussion of Fifteen Technical Issues listed in Attachment G.
November 3,1976 Memorandum from Director NRR to NRR Staff", NUREG-0138, November 1976.
7.
" Pressure Mitigating System Transient Analysis Results" prepared by Westinghouse for the Westinghouse user's group on reactor coolant system overpressurization, dated July 1977.
8.
PSEG (Librizzi) letter to NRC (Schwencer) dated September 27, 1979.
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3 APPENDIX A i
SELECTED ISSUES PROGRAM
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TECHNICAL EVALUATION OF-THE ELEC RICAL, INSTRUMENTATION, AND CONTROL CE!IGN ASPECTS OF rn~E LOW TEMPERATURE OVERPRESSLRE PRCTE;T:CN SYSTEM FCR THE SALEM NUCLEAR POWER PLANT, UN:T 1 by D. H. Laudenbach*
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'EG&G, Er.ergy 'aas.;rren s Grou::
DUPLICATE DOCUMENT Entire document previously entered into system under:
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No. of pages:
.w 7590-01 UNITED STATES _ NUCLEAR REGULATORY COMMISSION DOCKET NO._50-272 PUBLIC SERVI _CE EL_ ECTRIC AND GAS _ COMPANY, PHILADEL_PHIA ELECT _RIC COMPANY, DELMARVA POWER AND_LI_G_HT COMPANY, AND ATLANTIC CITY ELECTRIC COMPANY NOTICE OF__ ISSUANCE OF AMEN 0 MENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 24 to Facility Operating License No. DPR-70, issued to Public Service Electric.and Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees), which revised Technical Specifications for operation of the Salem Nuclear Generating Station, Unit No.1 (the facility) located in Salem County, New Jersey.
The amendment is effective as of the date of issuanca.
.The amendment incorporates Standard Radiological Technical Specificatiors governing operation and surveillance of the low temperature pressurizer overpressure protecti,on system.
The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules -and regulations.
The Commission has made appropriate findings as required by the Act and the Commission's i
-rules and regulations in l' CFR Chapter I, which are set forth in the license amendment.
Prior public notice of this amendment was not required since the amendment does not involve a significant hazards consideration.
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7590-01 1
4 4.
The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to' 10 CFR $51.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of this amendment.
I For further details with respect to t,his action, see (1) the application for amendment dated June 29, 1978 as supplemented by letter J
dated September 27, 1979, (2) Amendment No. 24 to License No. DPR-70.
and (3) the Comission's related Safety Evaluation.
All of these it!ms 4
are available for public inspection at the Commission's Public Document i
. Room,1717 H Street, N.W., Washington, D. C. and at the Salem Free Public Library,112 West Broadway, Salem, New Jersey. A copy of items (2) and 4
(3) may be obtained upon request addressed to the U. S.
. clear Regulatory Commission, Washington, D. C.
20555, Attention:
Director, Division of 1
Operating Reactors.
Dated at Bethesda, Maryland, this 21 day of February,1980 l
FOR THE NUCLEAR P.EGULATORY COMMISSION
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f%2/d R A. Schwencer, Chief Operating Reactors Branch #1 l
Division of Operating Reactors-4
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