ML20141N900
| ML20141N900 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/07/1986 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8603180263 | |
| Download: ML20141N900 (8) | |
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Mr Northern States Power Company 414 Nicollet Mall
%nrieapons, Minnesota 55401 Telephone to12) 330 5500 t
March 7, 1986 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Request for NRC Reconsideration of Requirements for Pushbutton Covers and Annunciation of Containment Isolation and Containment Spray Reset The Safety Evaluation Report for Containment Purging and Venting During Normal Operation dated September 9, 1982 prepared by the NRC Staff required the installation of covers for the reset pushbuttons for containment isolation and containment spray and the installation of annunciators for these reset conditions. As discussed in the Northern States Power Company letter of August 23, 1982, these modifications were to be completed in conjunction with the detailed control room design review required by Supplement No, I to NUREG-0737.
The detailed control room review has now been completed. The review was summarized in a report submitted to the Commission by Northern States Power Company on December 31, 1985. As discussed on page 6-4 of this report, the completion of the above described modificatf.ons was found to be unnecessary and inconsistent with human engineering design practice.
The purpose of this letter is to request NRC Staff reconsideration of the requirement to install reset pushbutton covers and annunciators for containment isolation and containment spray reset.
Copies of the relevant sections of the NRC Staff Safety Ev11uation i
Report and our August 23, 1982, letter are attached. A description of the issues involved and our reasons for asking the NRC Staff to reconsider their position follows.
Description Based on the Detailed Control Room Design Review we ask that the NRC reconsider their requirement to annunciate the reset condition and put covers on the reset pushbuttons for containment spray and containment isolation.
D 8603180263 B6030/
00 PDR ADOCK 05000282 P
PDR i(
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Dir:ct:r cf NRR March 7, 1986 Page 2 1.
Containment Spray Automatic actuation of containment spray occurs when containment pressure reaches 23 psig and manual Reset occurs when it has dropped to 10 psig. Reset does not stop the pumps but does prevent automatic re-start unless the automatic initiating condition clears and returns. Manual start is always an option.
2-Containment Isolation Reset does not re-open containment penetrations but does allow opening penetrations individually if needed.
Reset does block automatic, but.not manual, re-initiation of containment isolation unless the Safety Injection Signal has been reset and re-initiated.
Covers on the reset pushbutton were to be used far administrative control and annunciation of the reset condition was apparently for "Think Again."
Reason for Request
In April, 1984 the new plant specific Emergency Operating Procedures (EOP's) based on the generic Westinghouse Owners Group Emergency Response Guidelines were put in use.
These procedures are effective administrative controls to prevent an inadvertent reset of either
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containment spray or containment isolation. The procedures also require continuous re-evaluation of the need for spray or isolation (an independent review function currently performed by the Shift Technical Advisor reviewing Status Trees, and the operators reviewing Red Path Summaries).
The procedures are far more effective as operator aids and administrative controls then either the covers or annunciators would be.
During the Detailed Control Design Review both of the above NRC requirements were considered at length by the Control Room Design Review Committee and it was unanimously concluded that based on the E0P's, on control board convention, and on human factors considerations neither of the modifications are appropriate for the control board (DCRDR Summary Report Section 6.1.4, attached).
In light of the review described above, we request that the NRC Staff reconsider the requirements for containment isolation and containment spray reset pushbutton covers and reset annunciation.
Please contact us if you require additional information related to our request.
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David Musolf Manager Nuclear Support Services Attachments c: G Charnoff NRC Resident Manager NRC Project Manager, NRC j
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UNITED STATES Director of NRR
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NUCLEAR REGULATORY COMMISSION tarch 7, 1986
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Attachment (1)
%...... J SAFETY EVALUATION REPORT FOR CONTAINMENT PURGING 'AND VENTING DURING NORMAL-OPERATION OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
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INTRODUCTION A number of events have occurred over the past several years which directly relate to the practice of containment purging and venting during nonnal plant operation. These events have raised concerns relative to potential failures affecting the purge penetrations which could lead to degradation of the containment integrity, and, for PWRs, a degradation in ECCS perfor-mance. By letters dated November 28, 1978, October 29, 1979 and March 8, 1982'we requested licensees of operating reactors to respond to certain generic concerns about containment purging _or venting during normal plant operation. The concerns are as follows:
- (1) Events had occurred where licensees overrode or bypassed the safety actuation isolation signals to the containment isolation valves. These events were determined to be abnormal occurrences and were so characs terized in our report to Congress in January 1979.
(2) Recent licensirig reviews have required tests or analyses to show that containment purge or vent valves would shut without degrading contain-ment integrity during the dynamic loads of a design basis loss of coolant accident (DBA-LOCA).
(3) Licensees who elected to purge (or vent) the containment were requested to demonstrate that the containment purge (or vent) system design met the criteria outlined in our Standard Review Plan (SRP) 6.2.4 and the associated Branch Technical Position (BTP) CSB 6-4, which have effec-tively classed the purge and vent valves as " active" involving the operability assurance program of SRP 3.9.3.
'During the interim period of our review of these generic conerns, the licensee committed to keep and has maintained the isolation valves in the purge and vent system closed whenever the reactor is operated above cold shutdown.
This commitment is to remain in effect until we have completed our review of the long term generic concerns which is the subject of this safety evalua-tion for the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2.
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. II. DISCUSSION AND EVALUATION' By letters dated January 5, April 12. July 10, November 14, 1979; Harch 17, June 3, November 7,1980; May 6, December 3,1981 and April 30, 1982 the licensee. responded to our generic concerns of containment purge and venting at the Prairie Island Nuclear Generating Plant Unit Nos. I and 2.
The licensee's responses have either completed or have committed to complete by certain schedular dates modifications to the purge and vent systems in order to resolve our concerns. Our evaluations of these concerns are as follows.
A.
Manual Override of Safety Actuation Sional Instances have been reported where isolation signals which are required to automatically close the purge and vent valves for achieving containment integrity were manually overridden to allow purging of the containment with a high radiation signal present. Consequently, we developed a position specifying that the design and use of all override circuitry be such that the Prairie Island Nuclear Generating Plant have protection needed during postulated accident conditions. The licensee responded to this concern by letters dated January 5 and April 12, 1979; March 17 and June 3, 1980 and May 6, 1981. As a result of the review of the licensee's submittals by our-consultant, EG8G, San Ramon Operations, the attached Technical Eval--
uation Report (EGG. San Ramon Operations Report No. 1183-4166 June 1981) provides their technical evaluation of the design complying with our criteria.
During the course of our review, the licensee committed to modifications that would remove the design capability to manually override the containment ventilation isolation actuation signal. These modification hava since been com31etedforbothunitsbythelicenseefAspartofthisreviewaction, K 7"The consultant also aucited tne oesign of other ESF systems against the same N
six criteria. The consultant determined that the use of the Safety Injection
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system reset will.not adversely affect other ESF systems such as Containment i
However, contrary to the report provided by our consultant, we have Spray.
i determined that the reset features directly associated with the other ESF systems (e.g. containment isolation rese't switches) do not fully comply with the NRC criteria. Specifically, we determined that the override / reset design
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for the Containment Isolations systems (Phase A and Phase B) and the Contain-I ment Spray system, do not conform to Criteria 2 and 3.
By letter dated
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i August 23, 1982, the licensee committed to modifications that will achieve
\\wconformance with these criteria.
In addition, by letter dated June 9, 1980 the licensee responded to our IE Bulletin No. 80-06 in which the safety system schematics were reviewed to assure that safety related equipment remain in the emergency mode after reset. Both units were further tested during refueling outages to verify that the as-built systems met the design criterion as described in the safety system schematics.
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- Conclusion Based on the above evaluation, our. review of our consultant's technical report, and plant modifications performed by the licensee, we conclude that the electrical control system at the Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2 meets our criteria for' averting the safety actuation signals Efran actu0 ting equipment associated with Containment Purge during reset modes. The d? sign of the " reset" circuits associated with the containment isolation systems ano containment spray system will be modified to ' achieve conformance with the criteria.
We therefore conclude that this matter is satisfactorily resolved._ We further find that the licensee has satisfactorly responded to our request in IE Bulletin No. 80-06 which we now consider complete.
This evaluation also satisfies our requirements concerning Item II.E.4.2,
' Position 4, " Design of Control Systems for Automatic Containment Isolation
- Valve" of MUREG-0737 TMI Action Plan.
B.' Containnent Purge and Vent Valves Ooerability During Design Basis Accioent Introduction -
By letters dated November 28, 1978 and October 29, 1979, we requested all licensees to provide test results or analyses to demonstrate the adequate capability of the purge isolation valves.to close against the dynamic forces of a design basis Loss of Coolant Accident (LOCA). The' licensee transmitted test results and analyses for the purge valves by letters dated June 5, November 14, 1979, December 3,1981 and April 30, 1982. These submittals tnclude a description of the purge systems, how the purge systems are used during plant operations and the analysis of the valve operability during accident conditions for the Prairie Island Nuclear Generating Plant Unit Nos. 1 and 2.
Discussion and Evaluation Two containment purge systems are installed in each unit at the Prairie Island Nuclear Generating Plant for containment purging and venting. The high volume purge and ventilation system (33,000 CFM) is used to ventilate containment following reactor shutdown to permit access for inspection and maintenance. Two 36 inch butterfly valves are provided on each supply and exhaust line. The licensee's submittal dated December 3,1981 indicates that the results of an analysis performed by the valve radufacturer showed these valves are not capable of withstanding LOCA-induced loads from the 4
full open position. Based on these results the licensee committed to keep the two 36 inch butterfly valves closed for both units for all operating modes above cold shutdown.
In addition, the licensee has also committed to install double gasketed blind flanges on the containment side of the penetrations of the large volume purge and vent system so that valve's resilient seals are not needed to perform an isolation function during plant operations above cold shutdown. We are requesting that the licensee submit f
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Dir:ctor of NRR March 7, 1986 Attachment (2)
Northem States Power. Company 414 Nicollet Mal Minneapoks.Menesota 55401 Teepnone (612) 330-5500 August 23, 1982 nirector Office of Muclear Reactor Regulation i
US Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GEhT, RATING PLA:TT Docket Nos 50-282 License Nos DPR-42 50-306 DPR-60 Centainment Purrine and Ventine during Normal Operation In a letter dated April 30, 1982 we provided additional information required by the NRC Staff and their contractors for the long-term review of the Prairie Island containment purge and vent system. We have also had followup telephone conference calls with our Project Manager in the Division of Licensing and other members of the Staff in recent weeks. During these discussions we informed the Staff of changes in the way we intend to implement modifications discussed in our April 30, 1982 letter and provided our co=ments related to a draft NkC Staff Safety Evaluation Report dealing with the contain=ent purge and vent system, purge and vent isolation reset logic, and reset circuitry used in other emergency safety feature logic. The purpose of this latter is to document the informat*.on we provided earlier on an informal basis.
36-Inch Containment Purge System Modifications Following installation of the blind flanges in the Unit 2 36-inch purge system as originally planned, it was found that the inboard valve shaf t seal could still provide a leakage path outside containment. To eliminate this concern, the modification to the 36-inch system will therefore also include removal of the inboard isolation valve intervals and actuator and plugging the shaft penetration with welded plugs or complete valve removal. The double gasketed seal will serve as the barrier to containment leakage inside containment. The blind flange will be removed only during cold and refueling shutdown conditions. A Type B test will be performed following replacement of the blind flange prior to going above cold shutdown.
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Y Director Office of NRR 8/23/82 Page 2 18-inch Containment Purge Svstem Modifications A recent review of the 18-inch containment purge modifiestion resulted in an alteration of our original plans. As originally designed, personnel would need to enter certainment to remove the blind flanges and install the debris screen. To reduce radiation exposure and increase safety..the plan now is to install the blind flanges outside containment, in the annulus, between the two existing isolation valves.
The debris screen will be permanently installed in the containment.
Refer to the attached figure.
-Administrative Controls and Annunciation of Logic Reset Reset of the containment spray and containment isolation logic are not currently annunciated in the control room. In addition, physical features are not provided to facilitate administrative controls of these reset features. During a telephone conference call on August 10, 1982 we agreed to provide annunciation of these reset conditions and provide ~ covers for the reset push buttons. These modifications to the control room board will be designed and implemented as part of the overall control room modification resulting from the human factor studies for Prairie Island. Modifications are currently planned for implementation in 1984 Technical Soecification Changes Within 90 days of receipt of the NRC Staff's Safety Evaluation Report related to this issue we will propose Technical Specification changes in the following areas:
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Type B tests for the blind flanges on both the 36-inch and 18-inch containment purge systems following replacement as prescribed by 10 CFR Part 50 Appendix J.
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Type C tests for the 18-inch containment purge valves prior to use of the system when reactor is above cold shutdown.
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The requirement to have blind flanges installed on the 36-inch purge system prior to going above cold shutdown.
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Director Office of MRR i
8/23 /82 Page 2 d.
The requirement to limit the use of the 18-inch inservice purge to "as low as reasonable achievable". tihen not in use, the blind flanges will be installed in the 18-inch systems.
Please contact us if you have any question
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David Musolf Manager of Nuclear Suppor. Services DMM/SAF/js Regional Admin-III, FRC Resident Inspector, NRC G Charnoff Attachment e
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Director of NRR March 7, 1986 Attachment (3)
Honeywell December 1985 DETAILED CONTROL ROOM DESIGN REVIEW
SUMMARY
REPORT PRAIRIE ISLAND NUCLEAR GENERATING PLANT (Report No. 10188-PI-8000)
Prepared for:
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Conmission Prepared by:
Northern States Power Company Nuclear Technical Services Department l
and Honeywell Inc.
Technology Strategy Center TECHNOLOGY SMARGY CENTER 1000 Boone Avenue N.
coiden valley,Mmnesota55427 Printed in U.S.A.
6-4 e Northern States Power Company has requested additional human f actors f
support from the Honeywell Inc., Technology Strategy Center throughout 1986.
6.1.4 Review of Nuclear Regulatory Commission Commitments control room' design review process is a useful mechanism for determining
'The resolutions to issues that arise during control room operating experience.
potential resolution to an NSP/NRC commitment to
.One such issue is a annunciate the reset conditions of containment spray and containment isolation logic and to provide covers for the reset push buttons (August. 23, 1982 letter from the manager of Nuclear Support Services at Northern States Power to the director of the Office of Nuclear Reactor Regulation).
The commitment stated that modifications necessary to provide this capability would be studied, designed, and implemented as part of the overall control room modifications
. resulting from the DCRDR.
The issues of annunciation of reset conditions of containment
- spray, containment isolation logic, and covers for reset push buttons were studied at -
the conclusion of the review phase of the DCRDR and also during the development of the control room instrumentation conventions specification--Prairie Island Human Engineering Design Eeouiranents and f
Cor.ver.tions Snecifications.
On the basis of the human engineering review, it was concluded that the lack of annunciation of reset conditions for containment spray and containment isolation logic does not constitute a human engineering discrepancy.
There appear to be no significant safety consequences attributable to the logic of the annunciation system for containment spray and contsianent isolation as i
i currently implemented.
On the basis of the evaluation of conventions specificatic ne for population stereotypes and good human engineering design practice, it was determined that covers for the push buttons used to reset containment spray and containment isolation were not a significant advantage over push buttons with no covers on the PINGP main control boards.
The lack of covers does tot constitute 'a human engineering discrepancy, and there appears to be no important safety rationale A
6-5 for covered reset push buttons.
Implementing such a modification would depart.
from conventional design at the plant and possibly contribute to some confusion about the reason for covers on some push buttons and not on others.
Our recommendation, according to the results of the control room design review I
regarding the.. annunciation...of re se t conditions for containment spray and containment isolation logic, and covers for reset push buttons, is that these modifications to 'the design are not warranted.
T 6.2 CORRECTION SCHEDULE 6.2.1 Coerection Implementation Categories The resolutions to the HEDs identified on the basid of the control room design
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review were categorized into the 12 groups shown in Table 6 -1.
These categories will be used to develop the modification plan.
6.2.2 Final Planned Corrections The final. set of planned resolutions for HEDs in priority categories 1, 2, 3, and 4 requiring correction is summarized in the following tables.
The summary tables are organized by implementation category in the following manner:
l e Panel A modification-Table 6-2 e Panel B modification--Table 6-3 e Panel D modification-Table 6-4 e Panel E modification--Table 6-5 e Panel F modification-Table 6-6 e Convention Specification enhancements--Table 6-7 Safety Parameter Display Systear-Table 6-8 e
e Plant Process Computer System--Table 6-8 Annunciator system redesign-Table 6-9 e
e Error Reduction Progran--Table 6-10 i
I e Regulatory Guide 1.97 commitments-Table 6-10 e Already corrected-Table 6-10
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