ML19323H913
| ML19323H913 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/20/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323H912 | List: |
| References | |
| NUDOCS 8006170253 | |
| Download: ML19323H913 (12) | |
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.uCLET NOS. 50-266 AND 50-3D1 1.0 Introduction By letter dated July 28,1977 (Reference 1) Wisconsin Electric Power Company (WEPCO) submitted to the NRC plant specific a'talyses in support of the reactor vessel overpressure mitigating system (OMS) for Point Beach Units 1 and 2.
The analyses were supplemented by letter dated October 28,1977 (Reference 2) and other documentation submitted by WEPC0 (References 3-6).
Staff review of all information submitted by WEPC0 in support of the proposed overpressure mitigating system is complete and has found that the systen provides adequate protection from overpressure tran-sients.
A detailed safety evaluation follows.
2.0 Background
Over the last few years, incidents identified as pressure transients have occurred in pressurized water reactors.
This term " pressure transients," as used in this report, refers to events during which the temperature pressure limits of the reactor vessel, as shown in the facility Technical Specifications, are exceeded.
All of these incidents occurred at relatively low temperature (less than 200 degrees 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 7) summarizes the technical considerations relevant to this matter, discusses the safety concerns and existino safety margins of operatinp reactors, and describes the regulatory actions taken to resolve this issue by reducing the likelihood of future pressure transient events at operatinp reactors.
A brief discussion is presented here.
J 8006i70
. 2.1 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, since reactor vessel steels are less tough and could possibly fail in a brittle manner if subjected to high pressures at low temperatures, power reactors have always operated with restrictions on the pressure allowed during startup and shutdown operations.
At operating temperatures, the pressure allowed by Apendix G limits is in excess of the setpoint of currently installed pressurizer code safety valves. However, most operating PWRs did not have pressure relief devices to prevent pressure transier.ts during cold conditions from exceeding the Appendix G limit.
2.2 Regulatory Actions By letter dated August 11, 1976, (Reference 8) the NRC requested that WEPCO begin efforts to design and install plant systems to mitigate the consequences of pressure transients at low temperatures.
It was also requested that operating procedures be examined and administrative
. changes be made to guard against initiating overpressure events.
It was felt by the staff that proper administrative controls were re-quired to assure safe operation for the period of time prior to instal-lation of the proposed overpressure nitigating hardware.
WEPC0 responded (References 5 and 6) with preliminary information des-cribing interim measures to prevent these transients along with some discussion of proposed hardware.
The proposed hardwdre change was to install a low pressure actuation setpoint on the pressurizer air operated rel'ief valves.
WEPC0 participated as a member of a Westinghouse user's group which was formed to support the analysis effort reauired to verify the adequacy of the proposed system to prevent overpressure transients. Using input data genera ted by the user's group, Westinghouse performed transient analyses (References 9 and 10) which are used as the basis for plant specific analysis.
Plant specif' analyses for Point Beach Units 1 and 2 were submitted by WEPC0 by letter dated July 18,1977 (Reference 1) and supplemented by letter dated October 28, 1977 (Reference 2).
2.3.1 Design Criteria Through this series of meetings and correspondence with PWR vendors and licensees, the staff developed a set of criteria for an acceptable overpressure mitigating system. The basic criterion is that the mitigating system will prevent reactor vessel pressures in excess of these allowed by Appendix G.
Specific criteria for system performance are:
l
. 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) Testabili ty: The system must be testable on a periodic basis con-sistent with tha System's employment.
- 4) Seismic and IEEE 279 Criteria:
Ideally, the system should meet seismic Category I and IEEf 779 criteria.
The basic objective is that the system should not be vulnerable to a common failure that would both initiate a pressure transient and disable the overpres-sure mitigating system.
Such events as loss of instrument air and loss of offsite power must be considered.
The staff also instructed the licensee to provide an alarm which nonitors 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.
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2.4 Design Basis Events The incidents that have occurred to date have been the result of operator errors or equipment failures.
Two varieties of.oressure transients can be identified: a mass input type fron charging pumps, safety injection pumps, safety injection accumulators; and a heat addition type which causes thermal expansion from sources such as steam generators or decay heat.
On Westinghouse designed plants, the most common cause of the over-l pressure transients to date has been isolation of the letdown path.
Letdown during low pressure operations is via a flowpath through the RHR system.
Thus, isolation of RHR can initiate a pressure transient if a charging pump is left running.
Although other transients occur with lower frequency, those which result in the most rapid pressure increases were identified by the staff for analysis.
The most limiting mass input transient identified by the staff is inadvertent injection by the largest safety injection pump.
The most limiting thermal ex-pansion transient is the start of a reactor coolant pump with a 50 degree F temperature difference between the water in the reactor vessel and the water in the steam generator.
Based on the historical record of overpressure transients and the imposition of more effective administrative controls, the staff believes that the limiting events identified above form an acceptable bases for analyses of the proposed overpressure mitigating system.
. 3.0 System Description and Evaluation WEPC0 adopted the " Reference Mitigating System" developed by Westinghouse and the user's group.
The licensee proposed to modify the actuation circuitry of the existing air operated pres-surizer relief valves to provide a low pressure setpoint at 425 psig during startup and shutdown conditions. When the reactor vessel is at low temperatures, with the low pressure setpoint selected, a pres-sure transient is terminated below the Appendix G limit by automatic opening of these relief valves. A wanual switch is used to enable and disable the low setpoint of each relief valve. The OMS will remain in service during heatup until the RCS temperature reaches a level corresponding to the value (approximately 370 degrees F) at which the inservice pressure test may be perfonned. Conversely during cooldown the OMS will be enabled when the RCS is depres-surized to a pressure less than 425 psig (the OMS setpoint) and before the RCS temperature drops below the temperature at which the inservice pressure test may be performed (%370 degrees F).
The staff finds the pressurizer relief valves with a manually enabled low pressure setpoint to be an acceptable concept for an overpressure mitigating system. Discus,sion and evaluation of the system proposed by WEPC0 follows.
3.1 Air Supply The power operated relief valves (PORVs) are spring-loaded-closed, air required to open valves, which are supplied by a control air source.
To assure operability of the valves upon loss of control air, a backup air supply is provided.
The backup air supply consists of a compressed gas bottle for each PORV.
Each tank contains enough air for approximately 139 valve openings. The staff finds the backup air supply to be acceptable.
3.2 Electrical Controls The PORV's are made operational for low-pressure reliuby utilizing a dual setpoint where the low-pressure circuit is energized and de-energized, depending en plant conditions, by the operator with a key-lock switch.
The logic required for the low-pressure setpoint is in addition to the existing PORV actuation logic and will not interfere with existing automatic or manual actuation of the PORV's.
The relief valves en the RHR system are available for pressure relief whenever the RHR system is connected to the RCS.
The RHR system is normally connected to the RCS during plant conditions when overpressuri-zation events have been most prevalent, i.e., during low-temperature and low-pressure conditions.
The RHR-system relief valves can be con-sidered a diverse relief system at Point Beach because the RHR system isolation valves do not automatically isolate the RHR system during a press;.re transient, thereoy making the relief valves available through-out the transient.
. Durin; :lant cocidown and prior to the collapse of the steam butoie in the pressurite r; the operator acting under administrative,
procedure places the keylock switch in the " low pressure" position ".
and ccnnects the RHR system to the RCS.
An alan, will alert the oper-ator when the pressure is sufficiently low so thrt activation of the lew pressure setpoint circuit will not inadvertently open a PORV.
Placing tne keylock switch in the " low pressure" position blocks the aiar-indicating low pressure and enables the low-temperature high-pressure alarm.
Durin; plant neatup, the operating procedure will identify the plant conditions for which low-pressure protection is no longer needed.
The coerator places the keylock switch in the " normal" position, there:y returning control of the PORV's to the operating high-pressure concition and avoiding inadvertent opening of the PORV's.
Placing the keylock switch in the " normal" position rem;ves or blocks the low-temperature high-pressure alarms and enables the low-pressure opera-ion alert alarm.
We find the above design features acceptable.
3.3.
Testability Testability will be provided.
WEPC0 has stated that verification of operability of the OMS control system will be performed prior to entering water solid conditions.
PORV testing will be performed during each refueling outage.
Testing requirements will be incor-porated in the Technical Specifications as discussed in Section 4.2 of this evaluation.
3.4 Appendix G The Appendix G curve submitted by WEPC0 for purposes of overpres-sure transient analysis is based on 32 effective full power years irradiation. The zero degree heatup curve is allowed since most pressure transients occur during isothermal metal conditions.
Margins of 30 psia and 10 degrees F are included for possible instrument errors.
The staff finds that use of this curve is acceptable a,s a basis for overpressure mitigating system per-formance.
i
d 3.5 Setpoint Analysis The one loop version of LOFTRAN (Reference WCAP 7907) was used to perform the mass input analyses.
The four loop version was used for the heat input analysis. Both versions reauire some input modeling and initialization changes were reauired.
LOFTRAN is currently under review by the 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 it simply the sum of the overshoot magnitude and the setpoint magnitude.
The PORV setpoint is adjusted so that aiven the setpoint overshoot, the re-sultant pressure is still below that allowed by Aopendix G limits.
WEPCO presented the following Point Beach Units 1 and 2 plant charac-teristics to determine the pressure reached for the desian basis pressure transients:
SI Pump Flowrate Flowrate versus pressure used in generic analysis (Ref.10) 3 RCS Yolume 6,900 ft PORY Opening Time 2 sec S G Heat Transfer area 44,000 ft Relief Yalve setpoint 425 psig Westinghouse identified certain assumptions and input parameters as conservative with respect to the analysis.
These include one PORV
. assumed to fail, conservative heat transfer coefficients, conserva-tive modeling of stored energy of steam generators, and conservative interpolation schemes to obtain plant specific results from generic analyses.
The relief capacity of the RHR (990 gpm at 500 psig) has been con-servatively neglected.
In fact the relief capacity of the RHR will accommodate hypothesized mass injection from a single safety injec-tion pump.
It is a diverse as well as redundant subsystem.-
3.5.1 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 the licensee with a series of curves based on the LOFTRAN analysis of a generic plant design shich 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 Point Beach Units 1 and 2 plant parameters to obtain a conservative estimate of the PORY set-point overshoot.
WEPC0 has taken credit for volumetric expansion of the RCS with increasing RCS pressure. This is calculated to reduce the pressure overshoot (P max - P setpoint) to 74% of the value calculated assuming no expansion. The calculations are docu-mented in References 1 and 2.
The staff finds this method of analysis to be acceptable.
The calculated pressure overshoot assuming inadvertent mass addition from a single safety injection pump is less than or equal to 94.5 psi.
With an OMS low pressure setpoint of 425 psig, the highest predicted pressure for the worst case mass input is 519.5 psig.
It has been assumed that only one PORY opens.
No credit has been taken for the RHR system relief capacity. The 32 EFPY Appendix G limit of Units 1 and 2 at RCS temperatures greater than or equal to 109 F and 136 F respectively is greater than or equal to 520 psig. Use of an Appendix G curve applicable for less than 32 EFPY's would show additional con-servatism.
Hence the OMS setpoint is considered acceptable.
3.5.2 Heat Input Case Inadvertent startup of a reactor coolant pump with a primary to secondary temperature differential across the steam generator, and with the plant in a water solid condition, was selected as the limitina j
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, heat input case. For the heat input case, Westinghouse provided the licensee with a series of curves based on the LOFTRAN analysis of a generic plant design to determine the PORY setpoint overshoot as a function of RCS volume, steam generator UA, initial RCS tem-perature, reactor coolant / steam generator aT, relief valve setpoint, and relief valve opening time.
WEPCO calculated the following values of the maximum pressure for the heat input transient for a fixed 6T of 50 degrees F as a function of the initial RCS temperature.
RCS Temoerature Maximum Pressure 100 441 140 454 180 465 2E.
490 In all these cases, for the given RCS temperature, the Appendix G -
limits are not exceeded.
The staff finds that the analyses of the limiting mass input and heat input cases show a maximum pressure transient below that allowed by Appendix G limits and is therefore acceptable.
3.6 Implementation Unit 1 WEPC0 installed interim hardware protection (a single control systen, without redundant air supply) during the fall of 1976. Redundant control channels were installed during the October 1977 refueling.
Redundant air supplies were installed on Unit 1 during the ;lovember 1979 refueling outage.
Unit 2 WEPC0 installed interim hardware protection during the spring 1977 refueling.
Redundant control channels and backup air supplies were installed on Unit 2 during the fiarch 1979 refueling outaqe.
h
.- 4.0 Administrative Controls To supplement the hardware modifications and to limit the magnitude of postulated pressure transients to within the bounds of the analysis provided by the licensee, a defense in depth approach is adopted using procedural and administrative controls.
Those specific conditions required to assure that the plant is operated within the bounds of the analysis will be enumerated in forthcoming Technical Spccifications.
4.1 Procedures A number of provisions for prevention of pressure transients have been incorporated in the plant operating procedures.
With respect to hypothesized mass addition transients, HPSI pump, HPSI isolation valve motars, and accumulator isolation valve motors are electrically isolates with circuit beakers locked in the open position.
Of particular concern is the conduct of the loss of A.C. simultaneous with Safety Injection test which may be performed with the P.CS in a water sol.id condition. During this test which simulates loss of A.C.-
power and startup of emergency diesel generators both safety injectJon pumps and safety injection isolation valves receive command signals.
The OMS is not designed to mitigate mass addition from both safety injection trains.
Additional steps will be incorporated in the plant operating procedures to check.that Safety Injection Isol.ation Valves, M0V866A and B are closed, and circuit breakers providing power to the valve motor operators are open and tagped, prior to performing this test.
With respect to heat addition hypothesized tra'nsients, RCP starts are minimized, RCS/SG differential temperatures are checked' prior ~ to RCP starts.
In addition RCP starts during solid water conditions are performed with minimized charging flow rates and maximized letdown flow rates.
With respect to any scenario, additional, redundant, relief capacity is provided by insuring that the RHR system is aligned prior to taking the plant solid.
The staff finds that the procedural and administrative controls described are acceptable.
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4.2 Technical Specifications To assure proper operation of the overpressure mitigating system, WEPC0 submitted, by letter dated November 2,1978, proposed Technical Specifica-tions for Staff review. The proposed specifications were reviewed against the following criteria.
1.
The OMS is to be operable when the RCS temperature is below the value at which inservice pressure testing may be performed.
The OMS setpoint is to be incorporated in the Technical Specifi-cations.
Operability requires that the system is enabled, up-stream isolation valves open and backup air supply charged.
Should one redundant train (control circuitry and associated relief valve) to inoperable for more than seven days either a vapor bubble is to be established in the pressurizer or the primary system depressurized and vented to the atmosphere within eight hours.
Should both redundant trains be found inoperable either a vapor bubble is to be established in the pressurizer or,
the primary system depressurized within eight hours.
2.
Suitable surveillance requirements are to be proposed consistent with the need for use of the OMS.
3.
Electrical isolation of HPSI pump and isolation valve motors and the reinstation of electrical power to these components is to be incorporated in the Technical Specifications.
4.
Surveillance requirements consistent with the assumption that the RCS/SG differentia 1 temperature is less than or eoual to 50 degrees F are to be proposed.
It is noted that calculations based on a RCS/SG aT of 50 degrees F result in ample margins to the Appendix G curves.
Should WEPC0 chose to damonstrate that larger values of RCS/SG iT are acceptable with respect to violations of the Appendix G curves corresponding relaxation of surveillance requirements will be accepted.
5.
Operation of the OMS (PORV's and/or RHR relief valves) to relieve a pressure transient is to be reported.
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, Our review of WEPCO's submittal indicated that some modifications and additions were required to ensure compliance with the Staff's criteria.
These changes have been discussed with and agreed to by the WEPC0 Staff.
With the inclusion of these changes, the Staff finds the proposed Technical Specificaticns conform to our criteria and are, therefore, acceptable.
5.0 Conclusions The administrative controls and hardware changes proposed by Wisconsin Electric Power Company provide protection for Point Beach Units 1 and 2 from pressure transients at low temperatures by reducing the probability of initiation of a transient and by limiting the pressure of such a transient to below the limits set by Appendix G.
The staff finds that the overpressure mitigating system meets the criteria established by the NRC and is acceptable as a long term solution to the problem of over-pressure transients.
However, any future revisions of Appendix G limits for Point Beach Units 1 and 2 must be considered and the overpressure mitigating system setpoint adjusted accordingly with corresponding adjustments in the license.
E nvi rc. mental Consiceration
'a'e nave determinec that the amend ents do not authorize a chance in ef fluent types or total amounts nor an increase in power level and Having made will r.ot result in any significant environmental impact.
this cetermination, we have further concluded that the amendments involve 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 issu'ance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, (1) because the amendments do not involve a significant that:
increase in' the probability or consequences of accidents previously considered and do not involve a significant decrease in,a safety margin, the amendments do 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 conducted in compliance with the Commission's regulations and the issuance of th2se amendments,will not be inimical to the common defense and security or to the health and safety of the public.
Date:
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. REFERENrCS 1.
W.E. letter (Burstein) to NRC (Case), July 18, 1977.
2.
W.E. letter (Burstein) to NRC (Case), October 28, 1977.
3.
W.E. letter (Burstein) to NRC (Rusche), September 3,1976.
4.
W.E. letter (Burstein) to NRC (Rusche), October 14, 1976.
5.
W.E. letter (Burstein) to NRC (Rusche), March 2, 1977.
6.
W.E. letter (Burstein) to NRC (Rusche), April 18, 1977.
7.
" Staff Discussion of Fifteen Technical Issues listed in Attachment G November 3,1976 Memorandum from Director NRR to NRR Staff."
NUREG-0138, November 1976.
8.
NRC letter (Lear) to W.E.,
(Burstein), August '11,1976.
9.
" Pressure Mitigating System Transient Analysis Results" prepared by Westinghouse for the Westinghouse user's group on reactor coolant system overpressurization, Jul.y 1977.
10.
" Supplement to the July 1977 Report, Pressure liitigating Systems Transient Analysis Results," prepared by Westinghouse for the Westinghouse user's group on reactor coolant system overpressurization, September 1977.
11.
W.E. letter (Burstein) to NRC (Denton), November 2, 1978 1
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