ML19353A856
| ML19353A856 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/07/1989 |
| From: | Carr K NRC COMMISSION (OCM) |
| To: | Risacher B HARFORD COUNTY, MD |
| Shared Package | |
| ML19324C302 | List: |
| References | |
| GL-88-20, GL-89-16, NUDOCS 8911160101 | |
| Download: ML19353A856 (4) | |
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UNITED STATES
- NUCLEAR REGULATORY COMMISSION e-J.
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-November 7, 1989 CHAIRMAN Ms. Barbara Ahern Risacher Councilwoman, District A County Council of Harford County 20 West Courtland Street Bel Air, Maryland 21014
Dear Ms. Risacher:
.I am responding to your letter of August 28, 1989, concerning the Nuclear.
Regulatory Commission's (NRC's), disposition of the recomendations in SECY-89-017. " Mark. I Containment Performance: Improvement Program," and the
-implications of our decision on containment improvements planned for the Peach l
Bottom Station, Units 2 and 3.
The Comission, af ter carefully considering SECY-89-017, decided on a different course of action to ensure that the recomended improvements will be implemented, where justified, after due consideration has been given to plant-specific design differences.
In its January 23, 1989 paper, the staff recommended five improvements for Mark I containment plants:
(1) improved hardened vent capabilit q
water supply to the reactor vessel and drywell sprays,-(4 reactor pressure vesselfdepressurization system reliabilit an alternate extended emergency
-procedures and-training, and (5) accelerated staff actions to implement the Station Blackout Rule. After considering these recomendations, the Comission
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' directed the staff for the first item to follow an approach of approving
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installation of a hardened vent for licensees who, on their own initiative, j
elect to incorporate this plant improvement. The staff also was directed to
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initiate plant-specific backfit analyses for the remaining plants to evaluate i
the efficacy of requiring the installation of such vents. This directive has'-
been comunicated to nuclear power plant licensees by Generic Letter 89-16,
" Installation of a. Hardened Wetwell Vent," dated September 1,1989 (Enclosure L
L 1). As stated in detail in the Generic Letter, the staff believes that the
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available information provides strong incentive for installation of a hardoned vent since it would contribute to improved accident management strategies and to a reduced likelihood of core melt.
Licensees, including the Philadelphia Electric Company, have been requested to respond within 45 days of receipt of a
.the letter with a description of their plans for addressing the resolution of this issue. The staff is currently reviewing Philadelphia Electric Company's response dated October 30, 1989 (Enclosure 3). The Comission has indicated 4,
its intent that the entire process, including implementation, be completed within three years.
E With regard to public participation, the staff plans, in a related effort, to prepare a generic environmental assessment of containment venting using the improved hardware and procedures. However, with the environmental assessment incomplete, decisions concerning the appropriate staff actions and degree of public participation in this process have not yet been made.
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.. w 2-Regarding issues (2) through (4), the Comission believed that with the additional design-specific insights that will be gained from the Individual Plant Examination (IPE) program, licensees and the NRC staff will be in a better position to assess for each plant the risk significance of these issues relative to other identified plant vulnerabilities. Accordingly, the Comission concluded that issues (2) through (4) should be evaluated as part of the IPE program.
Generic Letter 88-20. " Individual Plant Examination For -
Severe Accident Vulnerabilities," dated November 23, 1988, has been issued to request licensees to perfonn a systematic evaluation, defined as individual plant examinations, to identify and report any plant-specific vulnerabilities to severe accidents.
Generic Letter 88-20 was augmented by a-supplement dated August 29, 1989, to provide, among other things, a summary of the staff's conclusions and recomendations on issues (2) through (4) for consideration in each Mark I licensee's IPE (Enclosure 3).
With respect to the fifth issue, since the Station' Blackout Rule was issued in the Comission's regulations (10 CFR 50.63) on July 21, 1988, all nuclear power plant licensees have submitted information regarding their implementation of the rule. The staff has developed a station blackout review process for all plants-that reflects our objective of reducing the overall risk of station blackout expeditiously. Plants with Mark I containments, including the Peach Bottom plants, are included in the higher priority group of plants to be reviewed..The staff expects to complete the station blackout review for the Peach Bottom plant in the spring of 1990.
By letter dated September 25, 1989, the Philadelphia Electric Company informed you of its plans to respond to NRC's initiatives for Mark I containment im-provements. The NRC will evaluate the utility's response to these initiatives and develop findings on the various aspects of the Peach Bottom containment performance.
I hope that this information will clarify the Comission's position on the Mark I Containment Performance Improvement program.
If you have any further questions, please contact me or Mr. William T.. Russell, Regional Administrator, NRC Region I Office, 475 Allendale Road, Kir.g of Prussia, Pennsylvania 19406.
i Sincerely, M s. b Kenneth M. Carr
Enclosures:
1.
Generic Letter 89-16, Installation of a Hardened Wetwell Vent 2.
Philadelphia Electric Company's response to Generic letter 89-16 1
3.
Generic Letter 88-20, Supplement 1, i
Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 5 50.54(f) l w
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ENCLOSURE 1 UMTEO STATES
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September 1. 1989 TO:-
ALL HOLDERS OF OPERATING LICENSES FOR NUCLEAR POWER REACTORS WITH MARK I CONTAINMENTS
SUBJECT:
INSTALLATION OF A HARDENED WETWELL VENT (GENERIC LETTER.89-16)
As a part of a comprehensive plan for closing severe accident issues, the staff undertook a program to determine if any actions should be taken, on.
a generic basis, to reduce the vulnerability of SWR Mark I containments to severe accident challenges. At the conclusion of the Mark I Containment Performance Improvement Program, the staff identified a number of plant modifications that substantially enhance the plants' capability to both prevent and mitigate the consequences of severe accidents. The improvements that were recoamended include (1) improved hardened wetwell vent capability.
- 2) improved reactor pressure vessel depressurization system reliability
- 3) an alternative water supply to the reactor vessel and drywell sprays,, and
- 4) updated emergency procedures and training.
The staff as part of that effort also evaluated various mechanisms for implementing of these plant improvements so that the licensee and the staff efforts would result in a coordinated coherent approach to resolution of severe accident issues in accordance with the Commission's severe accident policy.
After:considering-the proposed Mark I Containment Performance Program (described in SECY 89-017 January 1989), the Commission directed the staff l
to pursue Mark I enhancements on a plant-specific basis in order to account for possible unique design differences that may bear on the necessity and nature of specific safety improvements. Accordingly, the Comission concluded that the recommended safety improvements, with one exception, that is, hardened wetwt11 vent capability,(should be evaluated by licensees as nart of the Individuai Plant Examination IPE) Program. With regard to the recommended plant improvement dealing with hardened vent capability, the Commission, in recognition of the circumstances and benefits associated with this modification, has directed a different approach.
Specifically, the Connission has directed the staff to approve installation of a hardened vent under the provisions of 10 CFR 50.59 for licensees, who on their own initiative, elect to incorporate this plant improvement. The staff previously inspected the design of such a system that was installed by Boston Edison Company at the Pilgrim Nuclear Power Station.
The staff found the installed system and the associated l
Boston Edison Company's analysis acceptable.
l A copy of Boston Edison Company's description of the vent modification is i
enclosed for your information. For the remaining plants, the staff has been directed to initiate plant-specific backfit analyses for each of the Mark I plants to evaluate the efficacy of requiring the installation of hardened wetwell vents. Where the backfit analysis supports imposition of that requirement, liable hardened vent.the staff is directed to issue orders for modifications install a re 2 09010325-
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.i Generic 1.etter 89-16 September 1, 1989 The staff believes that the available information provides strong incentive for installation of a hardened vent. First, it is recognized that all affected l
plants have in place emergency procedures directing the operator to vent under certain circumstances (primarily to avoid exceeding the primary containment pressure limit) from the wetwell airspace. Thus, incorporation of a designated capability consistent with the oojectives of the emergency procedure guidelines is seen as a logical and prudent plant improvement. Continued reliance on pre-existing capability (non-pressure-bearing vent path) which i
may jeopardize access to vital plant areas or other equipment is an unnecessary complication that threatens accident management strategies. Second, implementation of reliable venting capability and procedures can reduce the likelihood of core melt from accident sequences involving loss of long-term decay heat removal by about a factor of 10. Reliable venting capability is also beneficial, depending on plant design and capabilities, in reducing the likelihood of core melt from other accident initiators, for example, station blackout and anticipated transients without scram. As a mitigation measure, a' reliable wetwell vent provides assurance of pressure relief through a path with significant scrubbing of fission products and can result in lower releases even for containment failure modes not associated with pressurization (i.e.
linermeltthrough).
Finally, a reliable hardened wetwell vent allows for co,nsideration of coordinated accident management strategies by providing design capability consistent with safety objectives. For the aforementioned reasons, the staff concludes that a plant modification is' highly desirable and a prudent engineering solution of issues surrounding complex and uncertain phenomena. Therefore, the staff strongly encourages licensees to implement requisite design changes, utilizing portions of existing systems to the greatest extent practical..under the provisions of 10 CFR 50.59.
As noted previously, for facilities not electing to voluntarily incorporate design changes, the Commission has directed the staff to perform plant-specific backfit analyses.
In an effort to most accurately reflect plant specificity, the staff herein requests that each licensee provide cost estimates for l
implementation of a hardened vent by pipe replacement, as described in SECY 89-017.
In addition, licensees are requested to indicate the incremental cost L
of installing an ac independent design in comparison to a design relying on availability of ac power.
In the absence of such information, the staff will use an estimate of $750,000. This estimate is based on modification of
. prevalent existing designs to bypass the standby gas treatment system ducting and includes piping, electrical design changes, and modifications to procedures i
and training.
The NRC staff requests that each licensee with a Mark I plant provide notification of its plans for addressing resolution of this issue.
If the licensee elects to voluntarily proceed with plant modifications, it should be so noted, aiong with an estimated schedule, and no further information is necessary. Otherwise, the NRC staff requests that the above cost information be provided.
In either event, it requests that each licensee respond within 45 days of receipt of this letter.
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September 1, 15
~ This request is covered by Office of Management and. Budget Clearance Number 3150-0011, which expires December 31, 1989..The estimated average burden hours are 100 person hours per licensee response, including searching data sources, gathering and analyzing.the data, and preparing the required letters. These estimated average burden hours pertain only-to the identified response-related matters and do not include the time for actual implementation of the requested actions.
Send comments regarding this burden estimate or any ;
other aspect of-this collection of information, including suggestions for reducing this burden, to the Record and Reports Management Branch, Division i.
of Information Support Services, Office of Information Resources Management, U.S. Nuclear Regulatory Comission, Washington, D.C.
20555; and to the Paperwork Reduction Project (3150-0011), Office of Management and Budget.
Washington, D.C. 20503.
If you have any questions regarding this matter, please contact the NRC Lead Project Manager, Mohan Thadani, at (301) 492-1427.
Sincerely, 1
James G. Partlow Associate Director for Projects Office of Nuclear Reactor Regulation
Enclosures:
1.
Description of Vent Modification at the Pilgrim Nuclear Power Station 2.
List of Most Recently Issued Generic Letters l
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Ralph Gl Bird Semot vice P'eMed - %C'ea' BEco 88-126 3-U. S. Nuclear Regulatory Commission "I""
Document Control Desk Mashington, DC 20555 License DPR-35
'uccket 50-293 REVISED INFORMATION REGARDING PILGRIM STATION SAFETY ENHANCEMENT PROGRAM
Dear Sir:
Enclosed is a description of a revised design for the Direct Torus Vent System (DTVS) that was described in the " Report on Pilgris Station Safety Enhancements" dated July 1,1987 and transmitted to the NRC with Mr. Bird's letter (BECo 87-111) to Mr. Varga dated July 8,1987. This revision supersedes in its entirety the Section 3.2 included in the July 1,1987 report.
On March 7, 1988 Boston Edison Company (SECo) personnel met with Dr. Murley, Mr.. Russell, and Dr. Thadant and provided a tour of SEP modifications and an o
informal presentation of the quantification of competing risks associated with venting the containment and conclusions drawn from these results. This presentation provided BECo the opportunity to respond to questions posed under Item 1 Section 3.2
" Installation of A Direct Torus Vent System (DTVS)" in Mr.
Varga's letter to Mr. Bird of August 21. 1987 " Initial Assessment of Pilgrim i
Safety Enhancement Program".
The material presented was made available to the resident inspector and was included as Attachment II in NRC Inspection Report
- 88-12, dated May 31. 1988.
As you are aware from plant inspections we have installed the OTVS piping and portions of related control wiring. Currently, the OTVS is isolated from the Standby Gas Treatment System ($8GTS) by blind flanges installed in place of i
Valve A0-5025 and the DTVS rupture disk.
This configuration was inspected by NRR in the performance of a technical review which focused on System.
Mechanical Design and Structural Design issues. The review took place on March 2-3, 1988 as documented in NRC Inspection Report #88-07, dated May 6,1988 and determined the installation configuration to be acceptable. He now plan to remove these blind flanges and proceed with installation of Valve A0-5025 and
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the DTVS rupture disk. We conclude the valve and rupture disk provide equivalent physical isolation of the OTVS piping from the $8GTS and
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appropriately ensure the operational integrity of the S8GTS under design basis accident conditions.
Following completion of this work, we will perform a local leak rate test to verify that Valve A0-5025 is acceptably leak tight using the same method previously utilized in testing the blind flange. He also plan to complete all remaining electrical work on the OTVS in accordance with the revised design.
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a BOSTON EDISON COMPANY Augusc 18, 1988 U.S. Nuclear Regulatory Commission-Page 2 i
On the basis' of the revised Section 3.2, we conclude that the OTVS design as described in the enclosure does not require any change to the Technical Specifications and that ce can proceed with installation without prior NRC approval.
Please feel free to contact t.'e or Mr. J. E. Howard, of my staff at (617) 649-8900 if you have any questions pertaining to the design details of the DTVS.
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Attachment:
Section 3.2 Revision 1 " Installation of A Direct Torus Vent System (OTVS)"
JEH/amm/2282 cc:
Mr. D. Mcdonald Project Manager Olvision of Reactor Projects I/II Office'of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Station P1-137 Mashington. 0.C.
20555 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident Inspector Pilgrim Nuclear Power Station v
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f Attachment to BECo Letter 88126.
j Section 3.2 Revision 1 " Installation Of A Direct Torus Vent Systes (DTVS)"
Pages 14,- 15,16,17. - 18.19,19A, 198 -
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3,2 IWtTALLATION OF A DIttc7 Tokut vtut tyttrW (Dfvt) 3,2,)
Ohisetive of Damian chanas This design change provides the ability for direct venting of the torus to the main stack. Containment venting is one core damage prevention strategy utt11 red in the thR Owners Group toergency Procedure Guidelines (EPGs) as previously approves by the htC and is requitte in plant-specific toergency Operating Procedures (t0Ps).
The torus vent line connecting the torus to the main stack will provide an alternate vent path for implementing E0p requirements and represents a significant improvement relative to outsting plant vent i
For $6 pst saturated steam conditions in the capability.
torus. apporoximately 11 decay heat can be vented.
3.2.2 Denian chanas Descristian This design change.(Figure 3.2-1) provides a diru t vent path from the torus to the main stack bypassing the Standby Gas Treatment System ($8Gil).
The bypass is an 8" line whose upstream end is connected to the pipe between primary containment isolation valves A0-5042 A &
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3.
The downstream end of the bypass is connected to the t
00" main stack line downstrena of StGT5 valves A0N-100 and A0N-112. An B* butterfly valve 'AO-50tl). which can be remotely operated from the main control room is added downstream of 4" valve A0-504tt. This valve acts as the primary containment outboard isolation valve for the direct torus vent line and will conform to NRC i
recuirements for sealed closed isolation valves as i
defined in NUREG 0800 SRP 6.2.4.
The new pipe is A5ME 111 Class t up to and inclusive of valve A0 5025. Test connections are provided upstress and downstream of A0-5025.
The design change replaces the entsting AC solenoid valve i
for A0-50423 with a DC solenoid valve (powered from i
essential 125 volt DC) to ensure operability without dependence on AC power. The new isolation valve.
A0-5025, is also provided with a DC solenoid powered from the redundant 125 voit DC source.
Both of these valves are normally closed and fall closed on loss of electrical and pneumatic power. One inch nitrogen lines are added
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to provide nitronen to valves A0-50420 and A0-S025. New t
valve A0-5025 wl'l be controlled by a remote manual key-locked control svttch. During normal operatton, power to the A0-5025 DC solenoid util also be disabled by removal of fuses in the wiring to the solenoid valve.
This satisfies NUREG 0800 SRP 6.2.4. Containment Isolation System acceptance criteria for a sealed closed barrier.
An additional fuse will be installed and remain in place to power valve status indication for A0-5025 in the main control room.
14 Rev. 1 (7/25/88)
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NUREG 0800. SRP 6.2.4. Ites !!.6.F allows the use o sealed closed barriers in place of automatic isolation valves.
Sealed closed barriers include blind flanges and sealed closed isolation valves which may be closed remote-manual valves.
$Rp 6.2.4 cal): for administrative control to assurs that sealed closed isolation valves cannot be inadvertently opened. This includes mechanical l
devices to seat or lock the valve closed, or to prevent power from being supplied to the valve operator.
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Consistent with SRP 6.2.4. valve A0 50tl will be a seal closed remote manual valve under administrative cont to assure that it cannot be 1%dvertently opened.
i Administrative control will be maintained by a key-locked i
remote manual control switch and a fuse removed to prevent power free being supplied to the valve operator.
In accordance with NUREG 0737. Itea !!.t.4.2.7 Position
- 6. A0-5025 will be sealed closed and verified as such at least every 31 days.
A 20" plee will replace the existing 20" diameter duct between SBGTS valves A0N-108. AON-l'I and the entsting to" pipe to the main stack.
The estating 20" diameter i
duct downstream of A0 5042A is shortened to allow fitup i
of the new vent line branch connection.
1 A rupture disk will be included in the 8" piping t
downstream of valve A0-5025. The rupture disk will provide a second leakage barrier. The rupture disk is designed to open belce containment design pressure, but l
will be intact up to pressures equal to or greater than l
those which cause an automatic containment isolation during W accident conditions.
The two Primary Containsent Isolation Valves (PCIVs)
A0-50425 and A0-5025 are placed in series with the rupture disk.
No single operator error in valve operation can activate the OTV$. The rupture disk has a rupture pressure above the automatic containment high pressure trip point.
Thus, the inboard pCIV (AO-50428) will receive an automatic isolation prior to disk rupture. The inboard PCIV (AO 50428) requires physical electrical jumper installation to open at primary containment pressure above the automatic high pressure trip point.
Valve A0-5025 will be closed whenever primary containment integrity is required and DC power to its solenoid control valve will be disconnected. Indication of valve position v111 be provided in the main control room even with the valve power removed. Use of the direct torus vent will be in accordance with approved EPG requirements and controlled by E0Ps in the same manner as other existing containment vent paths. Prior to opening the vent va'ves the StGT system will be shutdown and valves A0N-108 and A0N-112 (the outlet of StGT) placed in a
. closed posttion. Rev. 1 (7/25/88) l
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New 4" vent pipe (8'-HSS-44). including valve A0-5025 is
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safety related. Vent piping downstreas of A0-5025 including StGTS discharge p' ping to main stack, is also All safety related piping will be safety related.
supported as Class !._ Nitrogen piping is non-safety related and will be supported as Class !!/I.
i The interpretation of the Class !!/1 designation through this report is given below:
All Class !! items which have the potential to degrade the inte Class !!grity of a Class ! ites are analyzed.
Such items do not require dependable mechanical or 5.
electrical functionality during $$t. only that all of the following conditions prevail:
t 1.
The Class !! items create no missiles which impact unprotected Class ! items safety functions.
2.
The Class !! iten does not deform in a way which would degrade a Class ! itoa.
3.
If the Class !! iten fails, then the Class ! ites is L
protected against the full tapact of all missiles L
generated by the assumed failure of Class !! items.
All electrical portions of this design are safety related except for the indicating lights on the MINIC panel C904, the tie-ins to the annunciator, and interface with the i
plant computer.
3.2.3 Damien Chanes Evaluatian i
3.2.3.1 Svatomn/t'amapnentt Affected i
fat,tainment Ahmannherie central tviten (CAct)
The torus purge exhaust Ilne inboard isolation valve A0-50423 and the associated 8' pipe are the components of the CACS affected by the design modification. With incorporation of the subject modification, the CACS will depend on both essential AC (for valve A0 5042A) and l
essential DC (for A0-50428) to perform its purging function.
The new 8' torus vent line will be connected to i
existing 8" CACS piping between valves A0-50428 and A0-5042A.
i Rev. 1 (7/25/88)
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I Standhv Gan Treatment twnten (18C71)
The StGTS fan outlet valves (A0N-108 and A0N-112), doctwork from these valves to the 20" line leading to the main stack, and the 20" line leading to the main stack are the components of this system affected by the-i proposed change.
Valve A0N-108 is normally closed, fail-open.
Valve A0N-112 is normally closed, fail-closed.
and these valves are provided with essential DC power and local safety related air supplies.
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Primary cantainment Italattan tvaten (pcit)
Valve A0-50428 is affected by the change from AC to DC power for the solenoid and by i
3 replacement of the outsting air supply with nitrogen.
The addition of containment outboard i
l-isolation valve (AO-5025) will not affect the PC15.
t Primarv tentainment tvaten (Pet)
L Valve A0-5025 acts as the primary containment L
outboard isolation-valve for the direct torus I
vent line and v111 confors to NRC requirements for sealed closed isolation valves as defined in NUREG 0800 SRP 6.2.4.
3.2.3.2 1&fatv Functione c<f Affected tvatems/P=trents Cantainment Atmannheric central Svates This system has the safety function of reducing the possibility of an energy release within the primary containment from a Hydr 0 gen-Oxygen L
reaction following a postulated LOCA combined with degraded Core Standby Cooling System.-
Standhv Gan Treatment tvatam i
This system filters exhaust air from the reactor butiding and discharges the processed i
air to the main stack.
The system filters l
particulates and todines from the exhaust stream in order to reduce the level of airborne contamination released to the environs via the main stack. The $8GTS can also filter exhaust air from the drywell and the suppression pool, t
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-Primary cantainment Italation tvatem This system provides timely protection against the onset and consequences of design basis accidents involving the gross release of radioactive materials from the primary containment by initiating automatic isolation I
of appropriate pipe)ines which penetrate the i.
primary centainment whenever monitored i
l variables exceed pre-selected operational 11mits.
Primary cantainment tvaten i
e The primary containment system in conjunction l
with other safeguard features limits the release of fission products in the event of a l
postulated design basis accident so that offsite doses do not exceed the guideline values of 10 CFR 100.
3.2.3.3 Potential Effects en tafetv Functions i
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l containment Atmannherie central tvayam. Standhv
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'reatmant Stantaa. Primary canta< nannt Italation twntam and Primary cantainment tvatam l
The improvements change the A0-50428 solenoid l
l control from AC to DC enabling it to open (from its normally closed posttion) with no l
dependence on AC power availability. The l
entsting air supply to A0-50420 is being L
replaced by nitrogen.
Ductwork at the outlet of the StGTS is replaced with pipe and the new vent line is connected to
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the 20" line at the outlet of-the StGTS.
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Addition of a new 4" vent line with containment isolation valve A0-5025 off, the entsting torus i
vent inne could introduce a flow path under design basis conditions that could vent the l
containment directly to the stack bypassing the StGTS.
l 3.2.3.4 Analvain of Effects en tafety Functiann l
An analysis of the effects on the safety functions of CACS, StGTS. PCIS and PCS for the installation of the direct torus vent is l
described as follows:
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The change from AC to DC control and the L
replacements of air with nitrogen on A0-50428 does not adversely affect the ability to open A0-50428 when the containment is being purged, or to isolate under accident conditions.
13 Rev. 1 (7/25/88)
j The endifications to the ductuork and 20' line leading to the main stack do not affect the design basis safety function of any of the safety related systems.
During normal plant operations, the CACS and the StGTS do not use the torus 20* purge and vent 11ne to perform their safety functions.
The containment isolation valves are in their i
normally closed position..thus maintaining primary containment boundary integrity.
There are no adverse affects on the primary containment systen by the addition of the OTV5.
Valve A0-5025 vill conform to NRC i
criteria for sealed closed isolation valves as defined in NUREG 0000 SRP 6.2.4 and will not affect design basis a:cidents. Use of the OTVS will be in accordance with the containment venting provisions of EpGs as approved by the NRC and controlled by (ops in the same manner as other existing containment vent paths.
The effects on the torus of the new 8' piping and A0-5025 have been evaluated for Mark I program loadings, using ASME SPVC Section !!!
criteria. The remaining piping including the rupture disk was evaluated using AN$1831.1 requirements.
During plant startup and shutdown (non-omergency condition) when the purge and vent line is in use, valve A0-5025 remains closed.
In addition, the rupture disk' downstress of valve A0-5025 will provide a.
second positive means of preventing leakage and prevent direct release up to the stack during containment purge and vent at plant startup or shutdemn.
During containment high pressure conditions, the torus main exhaust line is automatically i
isolated by the pCIS. There is no change to the existing primary containment isolation system function for A0 5042A or A0 50428.
The sealed closed position of valve A0-5025 and the additional assurance added by the rupture disk downstress will prevent any unadvertent discharge up the stack for all design basis accident conditions.
3.2.3.5 Danten channe Evaluation 0 ev canctustans l
Installation of the DTVS does not adversely affect the safety functions of the CACS. 58GTS, PC15 or the integrity of primary containment or any other safety related systems.
I i Rev. 1 (7/25/88) b
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Use of the OTV5 will be in accordance with the containment venting provisions of (PGs as l
approved by the M C and controlled by tops in the same manner as other existing containment j
vent paths.
The DTV5 provides an leproved i
containment venting capability for decay heat removal which reduces potential onsite and l
offsite impacts relative to the entsting L
containment venting capability.
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.e LIST OF RECENTLY ISSUED GENERIC LETTERS l
Date of Subject issuance Issued To
)
INSTALLATION OF A HARDENED 09/01/89 ALL GE PLANTS WETWELL VENT i
LETTER 89-16)(GENERIC GENERIC LETTER 88-20 08/29/89 ALL LICENSEES HOLDING OPERATING SUPPLEMENT NO. 1 (INITIATION OF THE INDIVIDUAL LICENSES AND PLANT EXAMINATION FOR SEVERE CONSTRUCTION VULNERABILITIES 10CFR50.54(f))
PERMITS FOR i
NUCLEAR POWER REACTOR FACILITIES EMERGENCY RESPONSE DATA 08/21/89 ALL HOLDERS OF SYSTEM GENERIC LETTER NO.
OPERATING LICENSES 89-15 OR CONSTRUCTION PERMITS FOR NUCLEAR POWER PLANTS CORRECT ACCESSION NUMBER IS 8908220423 SUPPLEMENT 1 TO GENERIC 08/21/89 ALL LICENSEES OF LETTER 89-07, " POWER REACTOR OPERATING PLANTS,
-SAFEGUARDS CONTINGENCY APPLICANTS FOR PLANNING FOR SURFACE
. VEHICLE BOMBS" OPERATING LICENSES, AND HOLDERS OF CONSTRUCTION PERMITS LINE-ITEMS TECHNICAL SPECIFI- 08/21/89 ALL LICENSEES OF CATION IMPROVEMENT - REMOVAL OPERATING PLANTS, 0F 3.25 LIMIT ON EXTENDING APPLICANTS FOR SURVEILLANCE INTERVALS OPERATING LICENSES, (GENERIC LETTER 89-14)
AND HOLDERS OF CONSTRUCTION PERMITS GENERIC LETTER 89-13 7/18/89 LICENSEES TO ALL SERVICE WATER SYSTEMS POWER REACTORS PROBLEMS AFFECTING BWRS, PWRS, AND SAFETY-RELATED EQUIPMENT YENDORS IN ADDITION TO GENERAL CODES APPLICABLE TO GENERIC LETTERS
. GENERIC LETTER 89-12:
7/6/89 LICENSEES TO ALL OPERATOR LICENSING POWER REACTORS EXAMINATIONS BWRS, PWRS, AND 3
YENDORS IN ADDITION TO GENERAL CODES i
APPLICABLE T0 t
GENERIC LETTERS I
i
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[E NOV-06 '99 17122 :D:NJCLEE SERVICES.
TEL @ 1-210-640-6773 n013 PS2
-l GL 89-16 j
ENCLOSURE 2 PHILADELPHIA ELECTRIC COMPANY NUCLEAR ciROUP HEADQUARTERS 95545 CHtsTERBROOK BLVD.
WAYNE. PA 190e7 56et (ain) ses esto ensevviva vics passement.=vei.saa October 30. 1989 Docket Nos. 50-277 50-278 t
License Nos. DPR-44 i
OPR-56 U.S. Nuclear Regulatory Coanission Attn: Document Control Desk Washington, DC 20555 i
SU8 JECT: Peach Bottom Atomic Power Station. Units 2 and 3 I
Generic Letter 89-16. " Installation of a Hardened Wetwell Vent" 1'
Gentlemen:
i NRC Generic letter dated September 1,1989. requ(GL) 89-16. " Installation of a Hardened Wetwell Vent,"
ired Philadelphia Electric Company (PECo) to submit a response which provides notification of our plans for addressing the hardened wetw vent issue.
GL 89-16 directed licensees with Mark I plants to voluntarily with plant modifications and provide an estimatsd schedule in the response; pr 1
otherwise. provide cost estimates for implementation of a hardened vent by pi replacement for NRC staff use in performing plant-specific backfit analyses.
Our response, provided in the attacament. provides notification that we will proceed with plant modifications to insrove the current venting capabilities at Peac Bottom Atomic Power Station.
The estas11thment of criteria and schedule for assessing and implementing potential modifications is described in the response If you have any questions, or require additional infomation, please cont us.
Very truly yours.
_ -m
=
Attachment W. T. Russell. Administrator. Region I. USHRC cc:
T. P. Johnson. USHRC Senior Resident Inspector Plugpsqag 8
%.e,
= d de C bn m e.M w SUICES TEL 2h-640-6773.-
--m nat2 P03 l
a ATTACHMENT PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 RESPONSE TO GENERIC LETTER 89-16
"!NSTALLATION CF A HARDENED WETWELL VENY" l
I t
and the recently published NUREG-1150, entitled Severe for five U.S. Nuclear Power Plants.
NUREG-115') provides a state-of-the-art understanding of severe accident risk and also provides an update of the risk from WASH-1400. Both studies concluded that the risk of a severe accident at Pea is extremely low.
Changes in plant configuration and procedures, the evolution of Probabilistic Risk Assessment (PRA) mathodology, and an increated understa severe tecident phenomena have all contributed to a f actor of 30 decrease in total core damage frequency (C0F) from that in WASH-1400 to that in NUREG-1150.
the most dominant scenario.from WASH-1400, the loss of decay heat removal (T In fact, i
decreased three orders of magnitude due to a more realistic assessment of containm i
venting using existing equipment and successful injection following venting.
Peach Bottom has implemented Revision 3 of the BW containment.
Detailed emergency procedures exist for each of the nine identified on the environment, personnel, and equipment. vent paths and are prioriti j
criteria of a scrubbed release with little impact on pers hard-piped 6 inch Integrated Leak Rate Test The path capable of handling depressurization flo(w rates associated with decay This particular flow path originates from the wetwell airspace and discharges outs the reactor building, The emergency procedures regarding venting were used as the basis in NUREG-11 determining the probability of failing to successfully implement the recuirements for i
venting. Given the time procedures of 1 in 100 (.01) was use,d to represe,nt operator failure.and hardware available, a fa Operating Procedure Inspection (50-27?(8)/88 200)NRC review In addition, an extensive in August, 1988.
team concluded that PEco was capable of carrying out the provisions of the E0PsThe ins concerning primary containment venting using er sting equipment except under the special conditions associated with station blackout.
To further improve the current venting capabilities at Peach Bottom, Philadelphia Electric Company (PECo will proceed with plant modifications.
NUREG-CR-5225 Addendum)1 and SECY 89-017 inoicate the The analysis in potential from installation of a hardened vent *.s achieved in reducing even further theprobabilityofthepostulatedlossofdecayheatremovalscenario(TW) assuming little credit for existing venting capability.
clean steam vent) as the assessment basis when determining the risk reduction i
4 potential of modifications at Peach Bottom.
Owner's Grout, to develop generic design critoria for the hardened vent.PECo will anticipated that the design criteria will be available for NRC It is 1990.
portions of the Individual Plant Examination (IPE) for Peach Bottom and studies the possibility of systems interaction effccts between the vent and existing plant design. Evaluation of containment venting inpa conducted during the Peach Bottom IPE process. ct on teenarios other than TW will be
wee- :. nid.bitatt+ zeJKES TEL O:1-315-640-6772 8013 F04 o
PEco Response to GL 89-16 Attachment Page 2 Using the risk reduction potential as a measure assures Philadelphia Electric will a plant-specific basis of assessing the most effective modifi health and safety.a continued PECo oosition of providing and enhancing the protec This maintains outage (Reload 9) at each unit.The modifications will be implemented prior the fall of 1992 for Unit 2 and fall of 1993 for Unit 3.These outages are cur t
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ENCLOSURE 3
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UNITED STATES l'
NUCLEAR REGULATORY COMMISSION j
wat e oTow.o.c.rosos
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August 29, 1989 T0:
ALL LICENSEE 5 HOLDING OPERATING LICENSES AND CONSTRUCTION FOR NUCLEAR POWER REACTOR FACILITIES
SUBJECT:
INITIATION OF THE INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCID VULNERABILITIES-10 CFR 550.54(f) - GENERIC LETTER N0. 48-20,
$UPPLEMENT N0. 1 This letter announces the availability of NUREG-1335, *!ndividual Plant Examination: Submittal Guidance ' (enclosed) and initiation of the Individual Plant Examination (IPE) process.,
In accordance with Generic Letter No.
88-20,l Register notice announcing the availability of the enclo licensees are requested to submit within 60 days from the date of the t
Federa document, their proposed programs for completing their IPEs. The proposed programs should be submitted to the U.S. Nuclear Regulatory Cosmission, Document Control Desk, Washington, DC 20555, and should:
i 1.
Identify the method and approach selected for performing the IPE, 2.
Describe the method to be used if it has not been previously submitted for staff review (the description may be referenced), and 3.
Identify the milestones and schedules for performing the IPE and submit-tir.g the results to the NRC.
NUREG-1335 was published in draft form in January 1989 and issued for public comment. All comments received, including those made during the IPE Workshop on February 24 through March 2,198g, and staff responses to them may be found in Appendix C of NUREG-1335. Licensees may find it useful In preparing their initial responses to review two options discussed on the matters of internal flooding and submittal format in Appendix C, in response to cassents 5.1 and 11.3 respectively.
In accordance with a recent Cosmission decision on staff recommendations for enhancements to BWR Mark I plants the staff plans to communicate directly with each licensee who possesses a, Mark 1 plant on the matter of a hardened vent path. A susmary of the staff's conclusions and recossendations for other potential Mark I enhancements is given in the enclosure hereto, for consideration to each Mark 1 licensee's IPE. Additional information is contained in SECY 89-017, ' Mark 1 Containment Performance improvement Progras," dated January 23, 1989.
The staff expects to issue conclusions and reconnendations for all other plants and containment types in about 6 months for similar consideration in IPEs.
Regulatory Basis Generic Letter 88-20 was issued pursuant to 10 CFR 550.54(f). A copy of the 10 CFR 50.54(f) evaluation which justified issuance of Generic Letter 88-20
^8906300001"
I i
Generic Letter 88-20 A0 gust 29, 1985,
Supplement No. 1 i
is in the Pubite Document Room. This supplement does not change the scope of Generic Letter 88-20. Therefore, there is no additional burden associated with this letter, and an OM8 clearance number is not required.
Sincerel g
Je s G. Partlow As ociate Director for Projects 1
Office of Nuclear Reactor Regulation
Enclosures:
1.
NUREG-1335 " Individual PlantExamInation:
Submittal Guidance,"
August 1989 6
2.
Mark 1 Containment Performance Improvements 3.
List of Most Recently Issued Generic Letters l'
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i 3M03 Federal Register / Vd. 54. No,100 / Friday. September 1. tese / Notices NAT)DNAL SC FOUNDATION Purpose o eeting: To attend NW., Lower 14 vel of the Gelman Instrut.tions sterials Development Building. Washington DC.
Moount Panel and p de advice and Pon puernese esponesanes coereacTt Name:Comau on Equal recommend one concer6ing K-2:
John H. Fleck. Offloe of Nuclear Opportunties in nce and Meth. Scien and Technology Regularory Research. U.S. Nuclear Engineenng.
education.
Regulatory Commisalon. Washington.
Place: Nation ence Foundation.
Agenda:
review and evaluate DC 305&&. Telephone (301) 4es-as79.
1800 G Street, h Weahington. DC Instruction aterials Development Deled in neckvine. Maryland eine asth day 20550.
proposale a art of the selection of August.1sse.
Dates: October 19, 20.1980.
pmcusIw ards.
Timee/ Rooms:
tober 18:
Rouni loatng:W proposals For the Nuclear Regulatory Comauseion.
Subcommittee on roons with being revie intJude informauon of a R. Wayne Howton,
%,, pg,j,g7,y w,,gg g
Diesbilities 9:00 a -12:00 p.m Room Propnety dendalincluding nature, ogj,,f g,g,,,,,,
g ctober 18:Su ttee on aci e ch e e a es and 5
" * ' ' " ' ' " ' "I'*" "'" * *!
Minonties 1:30 p
- 30 p.m Room Personal 1 ation concerning "8""****
go, individua ociated with the October 19: F aunittee Meeting Proposals.
as matters are within 9:00 a.m.--4 00 p. Room H0.
exemptions and (6) of 5 U.S.C.
(Desw h 4
October 20: S ommittee on Women 552b(c). Go ent in the Sunshine 9:00 a.m.-12:00 Act.
h Appued Eg Corp.:
leeuenee of a Deskten Undw Type of Mee Open.
Deted: A as, spee.
10CPR
Contact:
Mary. Kohlerman, g,g,b,,,,
Execuuve Secre of the CEOSE.
- Commene, ogement office.
(IIcense No.
so-ot)
Notional Scien undauon. Room 635.
Telephone Num
.202-357-7006.
(M Doc. to-Filed M1-ao. 8.45 em) pi e f,$g,
.)
y,",
p', c(
Safety and Sal ards, has taken acdon g
wit [h activities to en age full participation W
g to ti for ou der of groups curre undernpresented in NUCLEAR REGULATORY tif e profoulonaland COMMISSION Albrecht Rose Coordinator National Coali to Stop Food' Minutes: May obtained from the
. Irradiation, dat art.h 23.1988, with Eaecuuve Se at the above indivleluel Plant Eneminetton respect to %e led Radiant Energy addmu Corporation (
).ne Petitioner Agenda:To ow progress by the Assescvt Nuclear Regulatory mquated that
--o Me subcommittees, come familiar with Commission.
lasmuted to su ee use d coalum.
successfullate tion programe, and to acTtoet initetion of the Individual Plant me I with the tor and other NSF
[
Examination for Sever, Accident Vulnemomun' N Director ofRoe of Nuclear M. Robeene Material Safet d Safeguards has Committee Mon atOficer' spesesAnyt his notice announces the determined to the Petition. We
""8 availability of NUREG-1335. " Individual maunsIw &l 1 am explained in Plant Examinauon: Submittal the " Director's sion under to Cm (3 Doc. eMonas H1* e o aml Culdance." and initiation of the 2J0s." (D hich is avallable for Individual Plant Examinauon (IPE) public inspecti the Commiselon's process. In accordance with Generic Public Docum Room. 2120 L Street latter No. 88-20 licensees are requested (Lower L4 vel).
Washington. DC to submit wuhin 60 days of this notice.
20555. A copy his deciolon willbe inattuctional le Development PanelMeeun9 their proposed programs for completing filed with the tary for the their IPEs. The proposed programs Comrninion's iew in acewdence The National nee Foundation should be submitted to the U.S. Nuclear with 10 CR ion 2Mc)of the eYlnotrueal ta$e hingIon.bos55 hi r e on wi a
g hate ab tm 22.1980,
- 1. identify the method and approach l n"c'jf
,dt e e le m from 8:30 a.m. t p.m.
eclected for performing the IPE'used if it
- 2. Desenbe the method to be unless the Co inston on its own Place: Nation ence Foundation.
1800 C. St. NW.
'sehington. DC 20550.
has not been previously submitted for modon insti a review of the deciolon with het time.
Room #1242.
staff review (the descrapuon may be Type of Meet : Closed Meeting, referenced) and Deted at Rock le. Maryland this 24th day Contact Pero Alice J. Moses.
- 3. ldentify the milestones and d August,1989.
National Scien
'oundation,1800 C. St.
ochedules for performing the IPE and For the Nucle egulatory Commiselon.
NW., Washing. DC 20550 submitting the results to the NRC, Guy A. Arlotto, instructional rials Development, A copy of the IPE submittal guidance DeputyDirector, e o/Nuc earMaterial Room 635-A e (202) 357-7066.
(NUREG-1335)is available for So/etyendso/cs Minutes: Ma e obtained from the inspection and/or copying in the NRC (m Doc. as-20a4 ed u t-as:a.e aml Contract perso at the above address.
Public Document Room. 2120 L Street owwo coot n
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l Mark I Containment Performance Improvements j
1 The NRC staff has identified certain containment performance improvements that would likely reduce the vulnerability of the Mark I containment to severe accident challenges (Ref.1 and 2). The Cussaission expects that licensees of Mark I plants will seriously consider these improvements during their Individual plant Examinations.
It should be noted that these improvements should be considered in addition to improvements 1
that stem from the evaluation and implementation of the hardened vent.
(a) Alternate Water Supply for Drywell Spray / Vessel Injection:
An important improvement would be to employ a backup or alternate supply of water and a pumping capability that is independent of 4
normal and emergency AC power. By connectin pressure residual heat removal system (RHR) g this source to the low system as well as to the i
existing drywell sprays, water could be delivered either into the reactor vessel or to the drywell, by use of an appropriate valving arrangement.
An alternate source of water injection into the reactor vessel would greatly reduce the likelihood of core melt due to station blackout or loss of long-term decay heat removal, as well as provide significant accident management capability.
Water for the drywell sprays would also provide significant mitigative capability to cool cors debris, to cool the containment steel shell to delay or prevent its failure, and scrub airborne particulate fission products from the atmosphere.
1 A review of some BWR Mark I facilities indicates that most plants have one or more diesel driven pumps which could be used to provide an alternate water supply. The flow rate using this backup water system may be significantly less than the design flow rate for drywell sprays. The potential benefits of modifying the spray headers to assure a spray were compared to having water run out of the spray nozzles. Fission product removal in the small crowded volume in which the sprays would be effective was judged to be small compared with the benefit of having a water pool on top of the core debris.
(b) Enhanced Reactor Pressure Vessel (RPV) Depressurization System Reliability:
The Automatic Depressurization System (ADS) consists of relief valves which can be manually operated to depressurize the reactor coolant system. Actuation of the ADS valves requires DC power and pneumatic
i i !
i supply.
In an extended station blackout after station batteries have been depleted, the ADS would not be available and the reactor would be re-pressurized. With enhanced RPV depressurization system reliability, i
depressurization of the reactor coolant system would have a greater t
degree of assurance. Together with a low pressure alternate source of water injection into the reactor vessel, the major benefit of enhanced RPV depressurization reliability would be to provide an i
additional source of core cooling which could significantly reduce the likelihood of high pressure severe accidents, such as from the short-term station blackout.
l l
Another important benefit is in the area of accident mitigation.
Reduced reactor pressure would greatly reduce the possibility of core debris being expelled under high pressure, given a core melt and failure of the reactor pressure vessel. Enhanced RPV depressurization system reliability would also delay containment failure and reduce t
the quantity and type of fission products ultimately released to the environment, in order to increase reliability of the RPV depressurization system, assurance of electrical power beyond the requirements of existing regulations may be necessary. Perfonnance of the cables needs to be reviewed for temperature capability during severe accidents as well as the capacity of the pneumatic supply.
(c)EmergencyProceduresandTraining:
NRC has recently reviewed and approved Revision 4 of the BWR Owners Group EPGs (General Electric Topical Report NE00-31331 BWR Owner's Group " Emergency Procedure Guidelines, Revision 4," March 1987).
Revision 4 to the BWR Owners Group EPG is a significant improvement over earlier versions in that they continue to be based on symptoms, they have been simplified, and all open items from previous versions have been resolved. The BWR EPGs extend well beyond the design bases and include many actions appropriate for severe accident management.
The improvement to EPGs is only as good as the plant-specific E0P implementation and the training that operators receive on use of the improved procedures.
The NRC staff encourages licensees to implement Revision 4 of the EPGs and recognize the need for proper implementation and training of operators.
1.
E. Claiborne et al., " Cost Analysis for Potential BWR Mark 1 Containment Improvements," Science and Engineering Associates Inc.,
NUREG/CR-5278, SEA 87-253-07-A:1, January 1989.
2.
Wagner, K. C. et al., "An Overview of BWR Mark 1 Containment Venting Implications, Addendum 1: An Evaluation of Potential Mark 1 Containment Improvements, NUREG/CR-5225 Addendum 1, July 1989.
J
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^
LIST OF RECENTLY ISSUED GENERIC LETTERS
\\
Generic Date of i
Letter No.
Sub.iect Issuance Is5Ued TO j
88-20 GENERIC LETTER 88-20 08/29/89 ALL LICENSEES SUPPLEMENT 1 SUPPLEMENT NO. 1 HOLDING OPERATING
(!NITIATION OF THE INDIVIDUAL LICENSES AND PLANT EXAMINATION FOR SEVERE CONSTRUCTION VULNERABILITIES 10CFRS0.54(f))
PERMITS FOR NUCLEAR POWER REACTOR FACILITIES 89-15 EMERGENCY RESPONSE DATA 08/21/89 ALL HOLDERS OF l
SYSTEM GENERIC LETTER N0.
OPERATING LICENSES 89-15 OR CONSTRUCTION r
PERMITS FOR NUCLEAR POWER PLANTS 89-07 SUPPLEMENT 1 TO GENERIC 08/21/89 ALL LICENSEES OF LETTER 89-07, " POWER REACTOR OPERATING PLANTS, SAFEGUARDS CONTINGENCY APPLICANTS FOR PLANNING FOR SURFACE OPERATING LICENSES, i
I VEHICLE BOMBS" AND HOLDERS OF CONSTRUCTION PERMITS 89-14 LINE-ITEMS TECHNICAL SPECIFI- 08/21/89 ALL LICENSEES OF i
CATION IMPROVEMENT - REMOVAL OPERATING PLANTS, OF 3.25 LIMIT ON EXTENDING APPLICANTS FOR SURVEILLANCE INTERVALS OPERATING LICENSES, (GINERICLETTER89-14)
AND HOLDERS OF CONSTRUCTION PERMITS 89-13 GENERIC LETTER 89-13 7/18/89 LICENSEES TO ALL SERVICE WATER SYSTEMS POWER REACTORS PROBLEMS AFFECTING BWRS. PWRS, AND SAFETY-RELATED EQUIPMENT VENDORS IN ADDITION TO GENERAL CODES APPLICABLE *iG GENERIC LETTERS 89-12 GENERIC LETTER 89-12:
7/6/89 LICENSEES TO ALL I
OPERATOR LICENSING POWER REACTORS l
EXAMINATIONS BWRS, PWRS, AND VENDORS IN ADDITION TO GENERAL CODES APPLICABLE TO GENERIC LETTERS I
89-11 GENERIC LETTER 89-11:
6/30/89 ALL BWR PLANTS &
RESOLUTION OF GENERIC ISSUE ALL LISTINGS 101 " BOILING WATER REACTOR APPLICABLE TO WATER LEVEL REDUNDANCY" GENERIC LETTERS &
VENDORS, ETC.
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