ML13015A641
| ML13015A641 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/19/2012 |
| From: | Doerflein L NRC Region 1 |
| To: | Samson Lee, John Monninger, Bhalchandra Vaidya Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2013-0010, IR-95-006 | |
| Download: ML13015A641 (16) | |
Text
Bickett, Brice From:
Sent:
To:
Subject:
Attachments:
Importance:
Doerflein, Lawrence Thursday, July 19, 2012 2:42 PM Lee, Samson; Vaidya, Bhalchandra; Monninger, John; Pelton, David; Russell, Andrea; Bickett, Brice Fitz TI-121 IR FitzPatrick IR 95-06.pdf High The IR that closed TI-121 at FitzPatrick was 50-333/95-06.
I've attached the relevant portion of the report.
The Accession and micro fiche numbers for the transmittal letter and inspection report are:
Letter - 9505030079; 83721: 095- 099 Report-9505030085; 83721: 100-116 If there is anything else need, please let me know.
- Regards, Larry R 45ý 1
N",~
April 18, 1995 Mr. Harry P. Salmon, Jr.
Resident Manager New York Power Authority James A. FitzPatrick Nuclear Power Plant Post Office Box 41 Lycoming, New York 13093
SUBJECT:
NOTICE OF VIOLATION (NRC REGION I INSPECTION NO.
50-333/95-06)
Dear Mr. Salmon:
This refers to the results of the routine resident safety inspection conducted by Messrs. W. Cook and R. Fernandes from February 12, 1995 to March 25, 1995 at the James A. FitzPatrick Nuclear Power Plant, Scriba, New York.
A summary of the inspection findings was presented to you and members of your staff at an exit meeting on April 12, 1995.
This inspection was directed toward areas important to public health and safety.
Areas examined during the inspection are described in the NRC Region I inspection report, which is enclosed with this letter.
Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews.
Performance by the plant staff during this inspection period was mixed.
NYPA's deliberate and cautious approach to verifying the extent of fuel assembly debris and potential foel pin damage demonstrated a safety conscious philosophy.
On tha other hand, a number of personnel performance errors that involved radiation protection requirements and surveillance testing procedural noncompliances indicated carelessness by members of your staff for administrative controls.
Several examples of procedural noncompliance were cited and are discussed in detail in the enclosed report.
We note that these events occurred during your 1994-95 refuel outage and that NYPA has experienced similar performance declines during previous planned outages.
Your continued strong management attention is warranted.
You are required to respond to this letter and should follow the instructions specified in the enclosed Notice of Violation (Notice) when preparing your response.
Jn your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence.
After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.
The response directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of
- 1980, Pub. L. No.
96.511.
9505030079 95041B PDR ADOCK 05000333 OFFICIAL RECORD COPY IE:O1 Q
'V
Harry P. Salmon, Jr.
2 Your cooperation with us is appreciated.
Sincerely, Original Signed by:
Curtis J. Cowgill, Chief Projects Branch No. I Division of Reactor Projects Docket No. 50-333
Enclosures:
- 2.
NRC Region I Inspection Report Number 50-333/95-06 cc w/encl:
S. Freeman, President R. Schoenberger, Chief Operating Officer W. Cahill, Jr., Executive Vice President and Chief Nuclear Officer W. Josiger, Vice President - Nuclear Operations J. Kelly, Vice President - Regulatory Affairs and Special Projects T. Dougherty, Vice President - Nuclear Engineering R. Deasy, Vice President - Appraisal and Compliance Services R.'Patch, Director - Quality Assurance G. Wilverding, Manager, Nuclear Safety Evaluation G. Goldstein, Assistant General Counsel C. Faison, Director, Nuclear Licensing Supervisor, Town of Scriba C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law Director, Energy & Water Division, Department of Public Service, State of New York State of New York, SLO Designee
U.S. NUCLEAR REGULATORY COMMISSION Region I Report No.:
Docket No.:
License No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
95-06 50-333 DPR-59 New York Power Authority James A. FitzPatrick Nuclear Power Plant Scriba, New York February 12, 1995 through March 25, 1995 W. Cook, Senior Resident Inspector R. Fernandes, Resident Inspector B. Korona, Reactor Engineer, DRS curtisJ cOwll, Actig' Chief" Reactor Proj ct Section IB, DRP INSPECTION
SUMMARY
Routine NRC resident inspection of plant operations, maintenance, engineering, and plant support.
RESUMLT:
See Executive Summary 9505030085 950418 PDR ADOCK 05000333 0
1PDR
TABLE OF CONTENTS
- i~i~i: :i:Page No,
,,,L 1.0
SUMMARY
OF FACILITY ACTIVITIES.......................
1 1.1 NYPA Activities.
1 1.2 NRC Activities 1
2.0 PLANT OPERATIONS (71707,93702,92901,62703).
I 2.1 Followup of Events Occurring During the Inspection Period.
2.1.1 Damaged Fuel Bundle During Refueling....
1 2.1.2 Control Room Emergency Ventilation Air Supply Fan..
2 2.1.3 Unexpected Carbon Dioxide (CO,)
Discharge During Testing 2.1.4 Reactor and Unit Startup
.4..............
4 2.2 Followup of Previously Identified Item......
4 3.0 MAINTENANCE (62703,61726,92902).
5 3.1 Maintenance Observation 5...
5 3.1.1 Irradiated Fuel Bundle Ispections 6
3.1.2 Primary Containment Integrated Leakage Rate Test...........
7 3.1.3 Residual Heat Removal (RHR)
Service Water Piping Leak 7
3.1.4 Diesel Tie Breaker Tripping.
8 3.2 Surveillance Observation 8
4.0 ENGINEERING (37551,92903,71707) 9 4.1 Relay Room and HVAC Control Logic Testing........
9 4.2 Engineered Safety Features System Walkdown................ 9 4.3 TI 2515/121 - Verification of Mark I Hardened Vent Modifications............
10 4.4 Previously Identified Items....
11 5.0 PLANT SUPPORT (71707,40500,92904)............
12 5.1 Radiological Controls J2 5.1.1 Escorted Visitor Not Issued a TLD..........
12 5.2 Previously Identified Items.................
13 6.0 MANAGEMENT MEETINGS (30702,71707).........
13 6.1 Exit Meetings..........
... 13 NOTE: The NRC inspection manual procedure or temporary instruction that was used as inspection guidance is listed for each applicable report section.
i
10ý Two component hangers on the A emergency filter train were not reflected on plant drawings and are a different-design than those similarly depicted on the drawings.
The licensee initiated drawing changes to reflect the supports and provided-th*e inspector with calculations to demonstrate the structural integrity offilter system supports.
The inspector reviewed the calculations and had no further questions.
Two MO4s, 70MOD-114 and 70MOD-113, were disconnected and lock wired in the closed and open position, respectively.
The inspector subsequently learned that the configuration was the result of a temporary modification (No.93-152) put into place to address single failure concerns previously identified by the licensee.
The temporary modification placed the modulatingjdampers in fail-safe positions to ensure that the control room is provided a maximum supply of fresh air and to ensure positive control room air pressure during emergency conditions.
The inspector also learned that a minor modification was in progress to convert this temporary~modification into a permanent system modification.
The inspector reviewed the minor modification package and had no further questions.
As part of the system walkdown the inspector identified to the NYPA staff that two MODs, MOD-113 andMOD-114, and flow element FE-102 were omitted from as-built drawing FB-35C, Rev. 12, Equipment Room Heating, Vent and.Air Conditioning.
However, the inspector noted that the components were identified on the control room flow diagram FB-45A and identified in the FSAR.
The inspector was concerned that despitt a temporary modification and a minor modification being processed for two safety-related components, this deficiency in the as-built drawing was not identified by the NYPA staff".
The inspector identified his concerns to NYPA.
NYPA's evaluation was not complete at the end of the period.
The issue will remain unresolved, pending completion of NYPA evaluation and subsequent NRC review.
(URI 95-06-03)
The inspectors concluded that the control room emergency ventilation system was operable, and with the exception of the observations noted above, had no further questions.
4.3 TI 2515/121 - Verification of Mark I Hardened Vent Modifications As part of a comprehensive plan for closing severe accident issues, the NRC staff undertook a program to determine if any actions should be taken, on a generic basis, to reduce the vulnerability of BWR 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 would substantially enhance the plants' capability to both prevent and mitigate the consequences of severe accidents.
Recommended improvements included improved hardened wetwell vent capability.
On September 1, 1989, the NRC issued Generic Letter (GL) 89-16, "Installation of a Hardened Wetwell Vent,* requesting licensees with Mark I containments to consider installation of hardened wetwell vent systems under the provisions of 10 CFR 50.59.
Using guidance provided in Temporary Instruction (TI) 2515/121, "Verification of Mark I Hardened Vent Modifications (GL 89-16)," the NRC is in
.......... t li
1*1 the process of conducting inspections to verify licensees' implementation of commitments made in response to GL.89B16.
By letters dated October 27, 1989, and July 25f, 1990, NYPA notified the NRC staff that it would defer its decisilon on hardened wetwell vent installation until the Fitzpatrick Indfvidual,Plant Examination (IPE) was completed.
The NRC reviewed the information provided'in those letters and also inspected the existing wetwell vent path at FitzPatrick.*
As a result of these activities, the NRC identified weaknesses in procedures and-operations training and also determined that all Boiling Water ReactO:rOwners Group (BWROG) criteria were not met by the licensee's current designh., However, because the design was expected to achieve the desired reductlion in,,,.core damage frequency, the NRC approved NYPA's request to defer -itsfdeciion to fully implement'the BWROG hardened vent general design criteria until completion of their IPE.
This approval and initial safety evaluation is documented in an NRC letter dated January 24, 1991.
The January 1991 letter also approved the deferral of improvements in operator training and procedures until IPE completion.
After the FitzPatrick IPE was completed, the NRC reviewed the changes-to the training and procedures proposed by the licensee.
The subsequent :NRC safety evaluation of the vent-related procedures and proposed training was transmitted by NRC letter dated April 27, 1992.
In a September 28, 1992 letter, the NRC determined that the current vent path met the intent of the BWROG design criteria, based on additional information S provided by NYPA and results of the previous NRC inspection of the vent path.
This letter forwarded the NRC staff's safety evaluation for the hardened vent.
In addition, the NRC found that the plant procedures and training were adequate to provide the information and guidance necessary for operators to effectively use the FitzPatrick hardened wetwell vent.
The inspector reviewed the referenced reports and determined that all applicable portions of TI 2515/121 for thel.hardened vent at FitzPatrick have been effectively addressed.
The FitzPatrick hardened wetwell vent met the intent of the NRC approved BWROG guidelines'.
Emergency Operating Procedures were available to direct the initiitidn and. termination of venting, operator training on the system was acceptable, and operators were found to be
-knowledgeable of the design and funqction of the system.
TI 2515/121 is closed.
.4.4 Previously Identified Items (Updated) Unresolved Item (92-14-01):* Relay Room CO, System Testing As noted in inspection report 50-333/95-02, Section 4.2.1, NYPA conducted special test procedure, STP-76AU, Relay Room Enclosure Integrity Test, to collect data for an engineering analysis performed in lieu of a full discharge test of the relay room CO, fire suppression system.
The engineering analysis was performed for NYPA by Yankee Engineering Services via a February 1995 Engineering Report to NYPA and captured under a memorandum (JAF-RPT-FPS-02009) dated March 2, 1995.
Based upon satisfactory completion of this engineering
Bickett, Brice From:
Sent:
To:
Subject:
Attachments:
Jennerich, Matthew Wednesday, March 14, 2012 7:22 AM Bickett, Brice SER Fitzpatrick Hardened Vent September 28, 1992 SER for Fitzpatrick Hardened Vent 28 Sept 1992.pdf 1
UNITED STATES NUCLEAR REGULATORY COMMISSION
,WASHINOTON.
- o. C.,2o65
,otcirt'r 28, 1992 Docket No, 50-333 Mr.
Ralph E. Beedle
[xecutivi! Vice President
- riuclear Generation POower Authority of, the SL;ILQe Of Nlew York 123 Main Street White Plains, New York 1OO0 Dear Mr.
Heedle:
SUI1JECT:
HARDENED WETWELL VENT CAPABILITY AT TlHE JAMES A. FIIZPAIRICK NUCLEAR POWER PLANT (FAC NOS.
M74868 AMD M82364)
As a part of a comprehensive plan for closing severe accident issues, the NRC staff undertook a program to determine if any actions should be taker, on a generic basis, to reduce the vulnerability of BWR Mark I containment,, to severe accident. chalienqes-At the conclusion of the Mark I Containment P'erformance Improvement Program, the NKC staff identified a number of plant modifications that substant ially enhance the plant's capability to both prevent andl mitigate the consequences of severe accidents.
One of the modifications recommended was improved hardened wetwell vent capability.
Af*tur considering the proposed Mark I Containment Performance Program (described in SECY 89-01+1, January 1939), the Commission directed the staff 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 improvemen.ts.
Accordingly, the Commission concluded that the recommended safety Improvements, with one exception, that is, hardened wetwell vent capability, should be evaluated by licensees as part of the Individual 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, directed the staff to facilitate in-tallation 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, On September 1,
- 1989, the staff issued Generic Letter 891"6, "Installation of a Hardened Wetwell Vent," wh~i.ch encouraged licensees to Implement a hardened wetwel I vent capability under the provisions of 10 CFR 50.59.
By letters dated October 21,
- 1989, and July 25, 1990, the Power Authority of the State of New York (PASNY) notified the NRC staff that it would defer making a decision on whether to install a hardened wetwell vent until the FitzPatrick Individual Plant Examination (IPE) was completed.
In those letters, PASNY-provided "plant specific" design information and engineering analyses that justified this approach on the hardened vent issue.
The NRC staff reviewed the information provided by PASNY in the stated letters, Additionally, on August 22, 1990, the staff inspected the existing wetwell vent path at the FitzPatrick plant.
As a result of the staff's review of PASNY's submittals, the inspection of the FitzPatrick wetwell vent path, and a PDR ADOCK 0o500033 R
Mr.
Ralph E. Beedle-,
- 2 -
SepMXT 28, 1992 review of the existing venting procedures and training, the NRC, by letter dated January 24, 1991, approved PASNlY's approach to defer its decision to fully implement the industry's hardened vent general design criteria until compiet. ion of the IPE.
By telter dated December 6, 1991, PASHY provided the NRC with Its final position regarding Jmpl eiintation of Lhe hardened vent design criteria along with ins ights gained from p)erform irg lhe IPE and the status of Investigations irtu accident management strategies asociated with severe accidents.
fn a letter datefd August 14,
- 1992, PASNY provided additional Information on the hardened vent capability.
PASNY determined that the current deslgn of the FitzPatrick hardened wetwell vent meets many of the Boiling Water Reactor Owners Group (BWROQ) design criteria and represents an acceptable deviation frow the remainder, Furthermore, PASNY concluded that hardware modifications needed to fully meet the BWROG design criteria are not necessary to ensure that the vent performs its decay heat removal and scrubbing functions and
ýiould nut produce significant public hlnriofits, Ila:,ed on thhe information provided by IASriY and the results of the inpectioni of the FitzPatrick hardene:d wv;twe II vent-path., the 11RC dt.-tormint.d that the cuTrrent vent. path ii i.,t't the hardened vent des or their initent.
Furtheriore, the NklC staff finds that the plant airid Irairning are adequate to provide Lh(
information and guidance lot, operators:
to effectively Luse the I itzPiatrick hardened wetwel)
.(:, pa,1ili.y.
therefore, t1he NRC sta.ff concludes that the existing capibil ity at the FitzPatrick plant i,. acceptable.
NRC staff has ign criteria procedures necessary vent wetwe I vsjnt A copy of' the staff's evaluatlon of the plant-specific features, procedures, and training related to the FitzPatrick hardened wetwel) vent capability is enclosed.
Ihis action completes our review activities assocIated with (L 89-16 and closes TAC Nos.
M74868 and M82364.
Sincerely,
//
~
St.even A. Varga, Director Division of Reactor Projects -
/Il Office of fluclear Reactor Regulation Safety Evaluation cc w/enclosure:
Se, next page M,
I 4FORO.
1 0
'1
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PS.9!_LA!_.1UQ.I LL_.II[ STATE O1 N..EW YORK
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(,*euric Letter (GL) 89-16 oncouraged licensees to implement a hardened. wetwell vent c;p;abilfty under the provision of 10 CFR 50.59.
By letter dated July 25, 1990, the Power Authorilty of the State,,
of flew York (PASNY, the licensee) slhmilI.tcd an analysis of the potential bonefits of a hardened wetwell vent at the James A. FitzPatrick ftuclear Puo.;u Plant (FitiPatri~ck).
lhe analysis indicated that the existing wetwell vent Is hardened and capable of wiLhstaniding anticipated vent ing pr, ssures, except for the interface with the standby gas treatment system (SGIS).
1ho SGTS Is located in a building adjacn(.t to the reactor bulldlnq.
PASNY affirmed Its willingness to make cost blneficial mo difications to fully iie,-L thp approved hardened vent qeneral (Jesijr (:rriteria; however, it wanted to defer such actions until comppleting Its individual plant examinal.ion (11'E) pro~jrain.
By letter dated January 24, 1991, the WRC staff approved the licensee's req(uest to Integrate thf: results of its IPE program into its decision to make any modifi.cations to the existing vent design to fully implement the approved hardened vent general design criteria.
Upon completion of the. IPE program, the licensee was to:
(1) provide the NRC with its final position regarding implementation of the hardened vent design criteria, and (2) use the results of the IPE to re-examine the venting procedures and-training of operators.
By letter dated December 6, 1991, the licensee provided this information along with Insights gained from performing the IPE and the status of investigations into accident management strategies associated with severe accidents.
In a letter dated August 14, 1992, the licurnsee provided additional information on the ha'dened vent capability.
2.0 £AVA1I The FitzPatrick plant has a hardened vent system that originates at the primary containment suppression chamber and terminates at the inlet to the SGTS.
The hardened vent system Is located in the reactor building while the SGTS Is located In a building adjacent to the reactor building.
lhe SGTS
- consists, In part, of a series of filters connected by sheet metal ducting with an oxpected rupture pressure of a few psig.
Outlet piping of the SGIS.Is routed through the building and to the;.plant stack.
The hardened vent piping Is rated for 150 pslg Internal pressure.
As the vent system is already hardened up to the SOJ S, the licensee performed an analysis to determine whether addtLional hardened piping should be added to bypass the SGTS and any 9210060336 920920 vDR ADUCK 0,3000333 IN PDR tW " '.
lll
..!.......I l additiona.1 modifications were necessary to meet the hardened, vent design criteria.
Through completion of the IPE, the licensee gained several Insights for post-accident venting, For the TW (loss of decay heat removal) accident. sequence, the containment pressure approaches the primary containment pressure limit (PCPL) of q4 psig i1n approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, The emergency operating procedures (COPs) then direct the operators to vent tý i containment:to maintain pressure below the PCPL.
If the containment is not vented, the pressure will continue to rise leading to failure due to overpressurization, The licensee calculated the core damage frequency (COF) with venting (1,92 E-G/yr) and without vent'ing (2.7? E-5/yr).
These calculations demonstrated a reduction In CDF by a factor of 14 due to venting.
For the station blackout (SBO) accident scenario, decay heat Is transferred to the suppression pool causing an increase in containment pressure.
Depletion of station batteries after about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> causes failure of the remaining core cooling systems and core damage ensues.
Core damage occurs approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> Into the scenario with containment pressure remaining below the PCPL vent.setpoint pressure of 44 psig, Therefore, the licensee has concluded that venting cannot be considered as a mitigative concept for an SBO event, under the guidance of the existing Emergency Operating Procedures.
During SBO sequences, core damage is calculated to occur around 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> whereas the
.pressure necessary to reach the primary containment pressure limit (PCPL) venting pressure occurs at approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
The January 24, 1991, NRC staff evaluation of plant-specific features, procedures, and training related to the hardened wetwell vent capability at the FitzPatrick plant concluded that the existing venting capability was expected to achieve the desired reduction in core damage frequency; however, the hardened vent path did not completely meet the hardened vent design criteria.
As a result, FitzPatrtck was allowed to Integrate the results of its IPE program into its decision to fully implement the hardened vent design criteria.
The following is an evaluation of the FitzPatrick position relative to the hardened vent design criteria.
Criterion (a):
The vent shall be sized such that under conditionsof:
(1) constant heat input at a rate equal t o I percent of rated-thermal power (unless lower limit justified by analysis),
and (2) containment pressure equal to the PCPL, the exhaust flow through the vent is sufficient to prevent the containment pressure from increasing, The FitzPatrick vent path will relieve pressure through parallel 6 and 12-inch lines.
Based on the licensee analysis, one percent decay heat (24.36 MW) produces 25, 13 Ibm/sec of steam at the PCPL of 44 psig or a volumetric rate of 269,964 it /sec.
Since the initial flow of gases through the vent will consist of nitrogen and steam, the licensee concluded that a conservative vent mass flow rate of 44.21 Ibm/sec was required to limit the primary containment pressure to the PCPL level.
The 6-inch lIne is capable of passing 17 Ibm/sec and the 12-inch line is capable of passing 11 ibm/sec, (Mn
u..
-3 Based on these results, FitzPatrick meets the vent criteria through usQ of the 12-inch line or combination of the.6 and 12-inch line.
The NRC staff concludps that criterion (a) has been met.
.Criterion (b):
The hardened vent shall be capable of operating up to the
- PCPL, It shall not compromise the existing containment design basis.
The PCPL at FitzPatrick is 44 psig. The hardened vent piping has a design pressure rating of 150 psig, with the exception of the SGTS which is located in a building adjacent to the reactor building.
The SGTS room contains sheetnietal ductwork and Filters which are assumed to fail under most venting scenarlos, After ductwork failure, high pressure venting will pressurize the SGIS ruom until failure of the access doors to the outside, They are double doors that normally open to the environment thereby providing a large release path for the steam mikture.
As a result, the pressurization on the reactor building wall will be limited to relatively low pressures which will be well within the wall structural capability.
Although failure of the sheetmnotal ductwork will] render the SGTS Inoperable, this failure should not affect any safety equipment located within the reactor building.
the S615 building is adequately isolated from the systems within the r*actor building by the reactor building wall.
Further, the containment design pressure is 56 psig and the PCPL. is 44 psig, Both values are well below the piping design pressure of 150 psig.
The NRC staff concludes that criterion (b) has been net.
Criterion (c):
The hardened vent shall he designed to operate during conditions associated with the TW sequence.
The need For SBO venting will be
-addressed cijring the IP[.
The FltzPatric.k hardened vent Is capable of relieving at least one percent of rated thernial power and withstanding the associated pressures, with the exception of the SGTS piping which Is asuvfned.to fail,
'The containment isolation valves in the vent path are also capable of operation at the PCPL,.
In the event electrical or pneumatic power is not available to operate the vent valves, manual operation from the reactor building is possible.
TheIPE determined that the PCI11 would be reached after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> into a TW sequence, which should provide sufficient time for any manual vent actuations, if required.
The PASNY also provided preliminary Insights into the need and feasibility of venting during SO0 sequences and was examining several new accident management strategies, However, since core damage would occur long before venting was needed, venting was. not credited in the IPE for an SBO event.
The NRC staff concludes that criterion (c) has been met.
Criterion (d):
The hardened vent shall include a means to prevent Inadvertent actuation, Inadvertent actuation of the hardened vent at FitzPatrick is prevented through several mechanisms.
The emergency operating procedures are specific as to when venting is to be perf'ormed, Venting involves operation of several valves
-4.
Frumn the relay room, which is physically separated from the control room, The IW sequence most likely would involve loss of some emergency power, and therefore, some manual vent valve operation would be required, Containment isolation skjnals from high drywell pressure and possibly high containment rad(lation would have to be bypassed.
Therefore, either the need for manual operition or deliberate bypass actions makes, the potential of inadvertent vent inrg a remoto possibility.
As a result, the NRC staff concludes that the intent of criterion (d) has been met.
Criterion (o):
The vent path up to and Including the second containment Isulation barrier shall be designed consistent with the design basis of the lplant.
- h, NRC staff concluded, in its January 24, 1991, evaluation of the hardened vent, design, that the vent path meets the design basis of the plant.
The NRC staff cOnE'iuries that criterion (e) has been met.
Criterion (f):
The hard vent path shall be capable of withstanding, without l*ss of functional capability, expected venting conditions associated with the TW sequence.
The NRC staff concluded, in its January 24, 1991, evaluation of the hardened vent design, that the vent piping, with the exception of the SGTS piping, was c*pable of withstanding, without loss of functional capability, all expected venting conditions, In addition, the NRC staff concluded that the damage to the SGIS may bQ an acceptable deviation pending completion of the IPE, The l icensee evaluated loss of the SGTS based on the IPE and performed a cost-henefit analysis for providing a hardened pipe bypass around the SGTS for SBO scenarios.
The licensee concludeO that loss of the SGTS was an acceptable cuinsequence of venting and that modifications to the pipin.9 configuration were not justified, Modifications to the piping configuration could reduce the offsite (lose but would not decrease the core damage frequency.
The NRC'staff concludes that the existing design is sufficient and that the intent of criterion (F) has been met, Criterion (g):
Radiation monitoring shall be provided to alert control room operators of radioactive releases during venting.
FitzPatrick will use the existing containment high range monitor (CHRM) and post.accIdent sampling system (PASS) to assess the radiological Consequences of
- ventInrg, These monitoring systems ara capable of assessing severe.accident conditions and will be operable under the environmental conditions associated with venting.
The CIIRM provide indication of radiation levels with the drywell.
The PASS can take samples from the drywefl,
- wetwell, suppression
- pool, and reactor coolant, The results from a PASS sample are available within the 3-hour criterion of.NUREG-0737.
The NRC staff concludes that the intent of critorion (g) has been met.
Criterion (h):
The hardened vent design shall ensure that no Ignition sources are present ii the pipeway.
- j r.
r In the January 24, 1991, evaluation, the N1RC staff Indicated that there was a potential for a hydrogen deflagration upon rupture of the SGTS ducts.
Large amounts of hydrogen could be produced during a core melt scenario; however, the TW sequence is prevented from progressing to a core melt by relieving both moss and energy through the containment vent.
Therefore, large amounts of hydrogen are not expected for the IW sequence.
- However, the EOPs are symptom based, not sequence based procedures, In the event that hydrogen is released into the SGTS room, the vent flow will also consist of nitrogen and steam which will provide some amount of natural Inerting.
In addition, the barrier br.,twenn Lhe SGTS room and the reactor bhilding is a Z-foot thick reinforced concrete wall which provides a barrier against the adverse consequences of a hydrogen deflagration.
A hard pipe bypass around the SGTS could prevent any hydrogen deflagration within the SGIS room, The liconsee estimated the cost of this modification at
$680,000.
The licensee concluded, that combustion in the existing vent path is not risk siqinificant and does not plan to modify the vent design, Based on th. uncertainty as to whether a combustible mixture could develop, the prevention p)otential of steam and nitrogen to suppress a hydrogen deflagratlon, the mitigation potential of the concrete wall between the SGTS roolm and the safety related equipment, and the costs associated with -
Ilo(jificat ions, the NRC staff concludes that the existing design is acceptable and the intent of criterion (h) has been met, As sLated'in the January 24, 1991, evaluation, the NRC staff identified several weaknesses in the technical and human factors aspects of F-AOP-35, "Post Accident Venting of the Primary Containment," which could prove detrimental to effective operator use of the procedure, Subsequent to the Issuance of that evaluation, F-AOP-35 was revised to provide significant improvements.including:
step clarification, more detailed instructions, unhanced caution statements, and standardized phraseology and format, Also noted in the January 24, 1991, evaluation wore several deficiencies In the operator training pertaining to containment venting.
Subsequently, the licensee has committed to integrate the results of the IPE into the operator tramining proriram.
This training will provide operators with guidance regardin9 severe accident phenomena such &s the consequences of venting during sv;vere accidents.
Other improvements to the operator' training program which have already been implemented Include:
- 1. Training which provided clarification of procedural references to the.FitzPatrick PCPL, containment failure pressure, and alternative methods of heat removal, and 2, Iraining which provided guidance on use of the 2" bypass line flowpath to protect the SGTS, unless flow is insufficient to counteract the decay heat addition to the containment thus requiring the main vent line to be used, lhe NRC staff has reviewed the revised venting procedure and enhancements to the operator, training as they relate to conformance to the human factor issues of
[.ho Standard Review Plan (NUREG-0800)
Sections 13,2.1, "Reactor Operator i,'rtiir,"," and 13.5,1, "Operating and Maintenance Procedures,"
The tJRC staff fijd,. the revised procedura) guidance and operator training acceptable.
Ih, I icenseu has identified several accident management strategies associated viit~h operation of the vent which may be beneflelal, These venting strateg.ie iitcitf,., vent ing until containment pressure Is reduced to near atmospheric p?;u.ure. and Initiatinq1 venting early for certain circumstincos.
Thi INRC sk.iff oigreus with the ) icensee's approach of bringing these issues to the al(,:-nt ion of the Bol ing Water Reactor Owners Group ({IWROG) for future generic co'..idn;ration.
- However, the NRC staff has concluded that the design-and p
oc,',.ur'es currf;ntly imnplemented at the FitzPatrick plant are sufflcient to
.,,;i ily the hýiridenod vent de(sign criteria and ensure adoqua te plant iafety.
.3U L{}C..L.U.i.QU fiw,..,I un the above evaluat ion, the 111R staff concludes that PASNY either meets Ih,:
,'tIuLwod vent deSi.in cr1turfa or i ts intent at the FitzPatrIck plant.
Ilnii*,'~lh,:r,*r the IIRC staff finds the revised 'procedural uuldance and operator
,.,iminy re(garding containment venting acceptable.
Therefore, the staff has
,ii*
t,.,':;*inlt thai. uxistinlq cuntainnient vent path capability at the FitzPatrick 1I.,111, Is!
,ICCC$)t able
.itI c.l, l Con tribtto rs M
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