ML111530441
| ML111530441 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/01/2011 |
| From: | Pilgrim Watch |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML111530440 | List: |
| References | |
| RAS 20407, 50-293-LR, ASLBP 06-848-02-LR | |
| Download: ML111530441 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. c. 20555 CHAaAMAN Mr. William R. Griffin Executive Secretary Town of Plymouth Office of the Selectmen 11 Lincoln Street June 21, 1990 Plymouth, Massachusetts 02360
Dear Mr. Griffin:
r 1 am responding to your letter of April 24, J990, cor.cerning the torus vent at the Pilgrim Nuclear Power Station. referred the 12 specific questions you raised in your letter to the Nuciear Regulatory Commission (NRC) stuff, and their detailed responSes are enclosed.
Some background information that may be nelpfu1 to you ;5 also enclosed.
1 hope the information we providing will lead to a bettel' understanding of generic issues associated with venting, and, in particular, how they relate to the Pilgrim Nucletir Power St&tion. If yau have any further questions, please contact me or Mr. T. T, Hartin, AdmInistrator of NRC's Region J Mr. Martin can be by telephone It (215) 337-5255.
Enclosures:
- 1. Background Information
- 2. Responses to Concerns 3, SECY-89-017 Sincerely, \1",
Kenneth M. Carr 4. Inspection Report No. 50-293/88-07
- 5.
Report No. 50-293/88-12 Enclosure 2 Response to Concerns Raised oy W.R. Griffin The following items briefly summarize current information concerning the hardened vent. They are organized as specific responses to issues raised in your letter to Chairman Carr. You should note that two descriptive terms routinely used within the industry mean the same thing: both the "direct torus vent" and the "hardened wetwell vent" describe the vent path to the stack. For purposes of the following responses, they are equivalent.
Question 1 (Q 1): What are the decontamination factors for the pool for various isotopes?
In other words, how well does the wet well pool scrub out the fission by-products, keeping the radioactive particles from releaSing to atmosphere?
Response:
Q 2: Response:
Except for the noble gases (consisting of the isotopes of Xenon and Krypton), which are not retained by the pool to any significant degree, the suppression pool is highly effective in scrubbing out and retaining particulate and volatile fission products.
Calculations as well as tests indicate the pression pool would be expected to have a re:alistic decontaminatior, factor (DF) for particulate and volatile:
fission products of about 100, depending upon the accident sequence and the temperature of the water. This means that about 1 percent of the particulate and volatile radioactivity entering the pool would be released to the atmosphere, and about 99 percent would be retained within the pool. The wetwell pool is highly effective with a OF of about 100 in scrubbing particulate and volatile fiSSion products, but not effective in scrubbing noble gases with a DF of 1. Please provide a graph of offslte radiation doses based on the possibility of a vacuum breaker valve remaining open at 10%, 25%, 50% and 100%. The staff does not have the off-site radiation dose evaluation requested in your letter. This type of failure was not considered in the design basis for the facility since it was not considered to be a credible event. The basis for the staff's position in this regard is as follows. The vacuum relief for both the drywell and wetwell ;s provided by blo 100 percent vacuum relief breakers located in two penetrations in the wetwell containment shell. These penetrations terminate in the reactor building, which is generally referred to as the secondary containment.
Q 3: Response: Each penetration consists of a vacuum breaker and an air operated butterfly valve in series. During normal operation.
both valves are closed; the vacuum breaker is maintained closed by the weight of the disk, and the butterfly valve is maintained closed by positive actuator air pressure.
In the event of a loss-of-coolant accident (LOCA), the increasing wetwell pressure will add to the closing pressure of the vacuum breaker. As a result, it is anticipated that during the entire positive pressure history within the containment, neither valve in the penetration will move from its closed position.
However. at the end of the pressurization phase. there is a potential for creating a negative pressure in containment.
This would occur only after the steam release from the reactor coolant system has ceased. As the wetwell pressure approaches atmospheric, the butterfly valve is opened, thereby allowing the vacuum breaker to properly function.
The vacuum breaker would begin to open when the wetwe11 pressure becomes 51 ightly sub-atmospheric.
Air fro!!! the reacte,r would restore the wetwell pressure back cO atmospheric.
The above sequence description has focused on the Design Basis .Accident (DBA). However. the sequence is equally valid for a large number of potential severe occident scenarios.
The ferer,ces would be I imitec! to the pressure rise rate and the max jmum pressu re and tempel'ilture va 1 ues reached curi n9 the event. TheSE: differences, however, "Iould not alter the events "s describec above. Thel-efore for purposes of consideration of vacuum breaker f"ilure, the staff's conclusions can be considered applicable for both DBA and severe accident events. Therefore.
during the entire positive pressure profile of the event. the penetration has two closed barriers in series. It is only during the end of the pressurization phase that the penetration is aligned into its vacuum breaker role. Because of this double barrier protection and the fact that both valves are not expected to change position during the pressurization phase of the event. the staff has concluded that failure of the penetration as a leak tight barrier ;s not credible and need not be considered in the design basis. The NRC has recommended venting at the containment deSign pressure as a minimum. or in the case of Pi1orim, at 60 psi. Why is the Pilgrim DrVS rupture disk at half that. at 30 psi? The fact that the Pilgrim DTVS rupture disk is designed to rupture at 30 psi is not related to the NRC's recommendation that specified the venting pressure at the containment design pressure.
The set pressure for the rupture disk does not control the venting pressure there are two closed isolation valves in the flow path. These two valves are normally closed and will be opened manually by the operator if venting is needed. Pilgrim's venting pressure in this case is consistent with the recommendations contained in Emergency Procedure Guidelines (EPG). Revision 4. These guidelines have been approved by the staff. The maximum containment pressure at which the operators are expected to open the vent valve is 56 pSig (not 60 psi). which is consistent with the NRC recommendation on venting pressure.
The rupture disk is deSigned to serve as an additional leakage barrier at pressures below 30 psi. It is designed to open below the containment design pressure, but will be intact up to a pressure equal to or greater than those pressures that cause an automatic containment isolation during any accident conditions.
Therefore, its presence in the line can effectively eliminate the negative consequences of inadvertent actuation of the vent valves at pressures below 30 psi. The set pressure of 30 for the rupture disk satisfies these design Q 4: What is the minimum containment pressure allowed by procedures at which the operators could open the DTVS outboard containment valve, AG-5025? P.esponse:
Q 5: Respor,se:
Usc uf the direct torus vent will be in accordance wi1:h approveC:
EPG requirements and will be controlled by Emergency Operatin, Procedures (EOPs). There is 110 spec i fi ed rr.i ni mum conta i nment pressure by the BWR Owners Group EPGs, Revision 4, at which the opera1:ors could open the DTVS outboard containment valve. There is a primary containment pressure limit (PCPl) of 56 pSig. Plant-specific supporting analyses are used to indicate when the operators should begin the venting procedure.
These analyses considered a number of plant parameters, including the pressure rise rate. These actions ensure that venting is used only if needed, that the conditions are beyond the design-basis-accident assumptions.
and that the pressures in the containment do not exceed the pePl limit. Please provide information on the reliability of the hydrogen and oxygen concentration monitors at Pilgrim. Hhat percentage of the time have both systems been accurately functioning?
The post-accident hydrogen/oxygen analyzers were in Janviiry 1985 as part of the post-Tf11 desi9n modificctions.
Since the installation.
one train (of two) was inoperable for blc days in November 1985, and one train was inoperct.-le for four days in ,January 1986. for a total of six days. At no time were both trains inoperable Simultaneously.
Technical Specification 3.7.A.7.c allows the reactor to operate for up to 7 days if one train is inoperable.
Q 6: Response:
Q 7: RE:spor.se:
Q 8: Response:
Q 9: Response: 1n addition, the containment atmospheric oxygen analyzer.
which monitors the oxygen concentration during normal operation, has been extremely reliaole.
The plant staff conservatively estimated this analyzer to have a reliability that exceeds 98 percent. Does the NRC concur that the use of the DTVS does not involve an unreviewed safety question?
Yes. As documented in NRC 1nspection Report No. 50-293/88-07.
dated May 6, 1988. the NRC inspected the installed DTVS design configuration and the licensee's evaluation and determined that they were acceptable.
Venting has been approved under previous versions of the EPGs. The direct torus vent is initiated by procedures under conditions specified by the EPGs. Because the outboard valve, AO-S025, is sealed closed and subject to leak te,ting, this valve satisfies the of 10 CFR Part 50, Appendices A and J, which are the regulation, for containment isolation and leak testing, respectivl<ly.
Therefore, the NRC concurred that the use of the DrVS does not involve an unreviewed safety question.
Does the NRC concur that
(;se of the OTVS does not I-equire changes to Pilgrim's Technical Specifications?
\'(:$, the NRC agrees that the use of the DTI'S does not requ i re changes tu Pilgrim's Technical Specifications.
Our inspection
- reports, were noted in the previous responses, incluaed consideration of po,sible TS changes, and we determined thilt none were needed. Does the NRC judge the DTVS to improve the safety at Pilgrim? Yes. The DTVS provides an improved containment venting capability for decay heat removal. The DrVS will prevent the majority of postulated los, of decay heat removal sequences from resulting in core melt and will mitigate the consequences of the residual sequences involving core melt where Venting through the suppression pool is found necessary.
Additional safety benefit, of DTVS are discussed in the previous background paragraphs.
Does the NRC conclude that the installation and use of the DTVS arE> acceptable under the of iO CFR 50.59? Yes. As 1'.'£ noted in '!:h*: response to Question 6, the staff inspected the design of DrVS at Pilgrim and found the installed system and aSSOCiated analysis acceptable.
Venting had been approved under previous versions of the EPGs. The direct torus vent is Q 10: Response: initiated by procedures under conditions specified by the EPGs. In addition, the installation or use of the direct torus vent will not increase the probability of a new accident.
Therefore, the installation and use of the DTVS are acceptable under the provisions of ID CFR 50.59. Furthermore, in a supplemental assessment of October 12. 1988, the NRC staff concluded that the Safety Enhancement Program (SEP) modifications being implemented in accordance with 10 CFR 50.59, including the DTV modification, would enhance the overall plant safety and performance of Pilgrim. Does the NRC conclude that Boston Edison has adequately considered the technical issues germane to the DTVS? Yes. Based on the noted inspections and reviews of the Pilgrim SEP, the NRC staff concludes that the safety issues associated with the DTVS have been adequately considered.
Q 11: Hhy was the automatic reclosure on high radiation of valve AO-5025 deleted during the design revision of the system? Response:
The reclosure of valve deleted because reclosure, if performed at high radiatioll levels, would isolate the vent flow path venting is needed tc mitigate the overpressure chal1er.ge.
Thus, automatic reclosure could dEfeat the purpose of the VEnt design. Q 12: GeneriC Letter 89-16 indicates some benefits of Q hardened well vent to reduce core damage frequencies during SSO [station blackout]
and ATWS [anticipated transient without scram] accident scenarios.
Is this true for Pilgrim? Response:
Yes. The isolation valves, AO-5025 and AO-5042B, are designed with ac independent power supplies.
These two valves are powered from essential dc power and are backed up with diverse nitrogen actuation capability.
Therefore, in case of an SBO event, the valves would be available for venting. The venting concept is mainly designed to slow overpressure transients of the ment. During some ATWS events, the pressure in the containment will rapidly increase.
Venting pressure could be reached in a matter of minutes rather than hours. Therefore, venting may not prevent contaiment failure because of the high containment pressurization rate but would provide additional time to scram reactor and delay the core melt.
Enclosure 1 Back round Information Related to Pil rim Station's lrect orus ent stem On January Z3. 1989. the NRC staff presented its recommendations on Mark I containment performance improvements and other safety enhancements to the Commission in SEeY 89-017. It represented the completion of the staff efforts on the Containment Performance Improvement (CPI) Program for Mark I containments.
The program was established to determine what actions, if any, should be taken to reduce the vulnerability of containments to severe-accident challenges.
From this point of view, the staff proposed that hardened vent capability would enhance plant with regard to both severe accident prevention and mitigation.
low probability scenarios in which multiple failures occur can lead to containment failure. Containment failure from these scenarios can result in a loss of cooling Vlater which is used to remove decay heat. The installation of a hardened vent greatly reduces the likelihood of early containment failure and, therefore, reduces tht: risks to the publiC because cooling capability is maintained.
For other for which core melt is predicted, ventin9 could be effective in delaying containmEr;t failure and in mitigating the release of fission products.
Although venting of the containment is currently included in BWR emergency operating procedures to improve the survivability of the containment, which acts as the last barrier for an uncontrolled release of radiation, it generally uses a vent path that inclUdes ductwork with a 10\' design pressure.
Venting under high-pressure severe-accident conditions coulc fail this ductwork, release the containment atmosphere into the reactor building, and damage equipment or contaminate equipment needed for accident recovery.
Venting through this ductwork may hamper Dr complicate post-accident recovery activities.
The installation of a reliable hardened wetwell vent allows for controlled venting through the wetwell while providing a path with significant scrubbing capability of fission products to the plant stack and prevents damage to equipment needed for accident recovery.
Based on the staff's recommendation, the Commission directed the staff to allow the licensees that elected to incorporate this plant improvement to install a hardened wetwell vent in accordance with the Commission's regulations (10 CFR 50.59). Plant specific backfit analyses were directed for the remaining plants with Mark I containments.
Where these analyses supported impOSition of a hardened vent, the staff was to issue orders requiring this modification.
Prior to the Com."ission decision in this mctter, numer()us diSCUSSions with butt-, industry groups and individual licensees were conducted.
These discussions inCluded meetings with Boston Edison (the licensee for Pilgrim).
The purpose of these discussions was to gather all available information relative to the hardened vent to enable the staff to an informed decision.
During this process, Boston Edison proposed to install the Direct Torus Vent System (DTVS). The licensee had concluded that it had sufficient information to commit to a specific design for hardened wetwell vents. The proposed modification was conSistent 111th the staff's generic finding for Mark I plants. However, the staff did not use the Pilgrim design as a test case, as is indicated in your letter.