ML20205H798
| ML20205H798 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/21/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205H785 | List: |
| References | |
| TASK-06-04, TASK-2.E.4.2, TASK-6-4, TASK-RR, TASK-TM NUDOCS 8601290218 | |
| Download: ML20205H798 (32) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
I wasHWGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO VENTING AND PURGING CONTAINMENT DURING PLANT OPERATION GPU NUCLEAR CORPORATION i
JERSEY CENTRAL POWER AND LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET N0. 50-219
1.0 INTRODUCTION
In the NRC staff's letter of November 29, 1978, the staff identified the generic concerns of purging and venting of containments to all operating reactor licensees and requested a response to these con-cerns. This is the staff's multi-plant generic action (MPA) B-24.
The staff's review of Jersey Central Power and Light Company's and GPU Nuclear Corporation's (the licensees) responses was interrupted by the j
TMI accident and its demand on staff resources.
Consequently, an Interim Position on containment purging and venting was transmitted to I
the licensee on October 23, 1979. The licensee was requested to imple-ment short-term corrective actions at Oyster Creek to remain in effect pending completion of our long-term review of its response to the staff's November 29, 1978 letter.
The staff and its contractors have been reviewing the licensees' responses 4
to the staff's November 1978 letter to close out the long-term review of this rather complex issue. The licensees' responses and the staff's letters are listed in Section 6.0, References. The components of this review are as follows:
l 1.
Conformance to Standard Review Plan (SRP) Section 6.2.4, Revision 1 and i
Branch Technical Position (BTP) C5B 6-4 Revision 1.
These documents were provided as enclosures to the staff's November 29, 1978 letter (Ref. 1).
2.
Valve Operability The staff required a program demonstrating operability of the valves i
in accordance with our " Guidelines for Demonstrative Operability of Purge and Vent Valves." These Guidelines were sent to the licensee in our letter of September 27,1979 (Ref. 3). There is an acceptable 1
alternative in lieu of completing the valve qualification program for the large butterfly-type valves. This would be the installation i
8601290218 860121 PDR ADOCK 05000219 P
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of a fully-qualified mini-purge system with valves 8-inch or smaller to bypass the larger valves. The system would meet BTP CSB 6-4 Item B.1.c.
3.
Safety Actuation Signal Override This involves the review of safety actuation signal circuits to ensure that overriding of one safety actuation signal does not also cause the bypass of any other-safety actuation signal.
4.
Containment Leakage Due to Seal Deterioration Position B.4 of the Branch Technical Position (BTP) CSB 6-4 presents that provisions should be made to test the availability of the isolation function and the leakage rate of the isolation valves in the vent and purge lines, individually, during reactor operations.
5.
TMI Action Plan Item II.E.4.2 Containment Isolation Dependability, TMI Item II.E.4.2, is related to the staff's generic concerns on purging and venting of containments.
The seven positions of Item II.E.4.2 are the following:
(1) that there be diversity in the parameters sensed for the initiation of contain-ment isolation (Standa d Review Plan (SRP) 6.2.4), (2 and 3) that all nonessential penetrations be automatically isolated by the containment isolation signal, (4) that resetting the containment isolation signal will not cause automatic reopening of containment isolation valves, (5) that the containment setpoint pressure that initiates containment isolation for nonessential penetrations is reduced to a minimum value compatible with normal operating conditions, (6) that containment purge valves that do not satisfy the operability criteria in BTP 6-4 in SRP 6.2.4 or the Interim Position of October 23, 1979, are sealed closed as defined in SRP 6.2.4 during operating conditions 1, 2, 3 and 4 and verified at least once every 31 days, and (7) that containment purge and vent isolation valves are closed on a high radiation signal.
6.
Systematic Evaluation Program (SEP) Topic VI-4, Containment Isolation System 10 CFR Part 50, General Design Criteria (GDC) 54, 55, 56 and 57, as implemented by SRP 6.2.4 and Regulatory Guides 1.11 and 1.141, requires isolation provisions for the lines penetrating the primary contain-ment to maintain an essentially leaktight barrier against the uncon-trolled release of radioactivity to the environment. This topic included (1) the electrical, instrumentation and control aspects of the override of the containment purge valve isolation and (2) the closure time, the use of resilient seals, the operability during accidents, etc., of the containment ventilation system isolation valves. These are also addressed in Section 4 and in numbers (1) to (4) in Section 5 above.
~
2.0 EVALUATION The status of our long-term review of the above items for Oyster Creek is as follows:
2.1 Conformance to Standard Review Plan (SRP) Section 6.2.4, Revision 1 and Branch Technical Position (BTP) CSB 6-4 Revision 1.
2.1.0 Introduction j
The staff reviewed the primary containment or drywell purge / vent system design and operating practices for Oyster Creek, and transmitted a request for additional information on this topic to the licensee by letter dated January 4, 1982 (Ref. 12). The licensee responded to the staff's request for information by letters dated July 15, 1982, January 13, 1983 and April 19, 1984 (References 13, 14 and 19).
2.1.1 System Description The purge / vent system at Oyster Creek utilizes containment isolation valves ranging in size from 2 inches to 18 inches in diameter.
See the figure on the next page and the list of valves in Section 2.2.2.
Also, the licensee has committed i
to replace the larger (greater than 2-inch diameter) isolation valves with valves which would meet staff operability requirements (Ref. 8). The staff's review is, therefore, based on the infonnation supplied by the licensee concerning these new valves.
Inerting/deinerting of the containment (drywell and torus) is performed using the larger valves in the inlet and exhaust penetrations in both the drywell and torus. The 2-inch valves are utilized for nitrogen makeup and for pressure control. The purge / vent system is configured such that the lines penetrating containment have two isolation valves in series.
2.1.2 Evaluation
~
The vent / purge system at Oyster Creek is not frequently used during plant operating) conditions above cold shutdown (about 100-200 hours per year, and is only used when necessary for the safe operation of the plant.
It is not relied on for temperature and humidity control inside the containment. The 2-inch lines are used for containment pressure control and for nitrogen makeup operations. The staff has concluded that the rationale used for the design and operation of the purge /
vent system at Oyster Creek, and the annual usage of the system, are acceptable for the following reasons:
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- 1) Operating experience at the Oyster Creek facility has shown that purge system operation has not been extensive (100-200 hours per year).
i 2)
Inerting/deinerting of the containment and pressure ^
control are necessary actions for the safe operation of i
the plant.
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- 3) Technical Specification 3.5.A.6 limits the use of the purge / vent system at Oyster Creek to within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch is in the run made i
(i.e., for inerting) and to within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to shutdown (i.e., for deinerting).
- 4) The cost involved in replenishing the nitrogen is an incentive for the licensee to restrict purging and venting operations.
The size of the purge and vent lines for Oyster Creek are 2",
4 8", 12" and 18".
The 20" lines to the Reactor Building atmosphere are the torus vacuum breakers with check valves to prevent air leaving containment and are not part of the containment purge and vent system for this review. The 8",
12" and 18" lines have butterfly valves and are part of the containment purge and vent system. Although the 12" and 18" are greater than the 8" in BTP CSB 6-4, this is acceptable because the plant is already built. The 2" line in the Nitrogen Makeup System is used for adding nitrogen to inert the containment and has air-operated valves and is not considered part of the containment purge and vent system for this review.
There is one line into the containment and two lines out: one from the drywell and one from the torus. The line from the i
torus is from the airspace separated from the drywell by water in the torus and by a drywell-torus vacuum breaker. This meets i
the position in BTP CSB 6-4 for one purge line and one vent line for the Oyster Creek drywell/ torus, Mark I containment.
The 2.5 to 5 second valve closure times for the new Clow valves listed in Section 2.2.2 are acceptable.
The valves close on manual initiation, reactor water level low-low setpoint and high drywell pressure. These are diverse signals. The valves can be closed by either off-site power or the emergency diesel power. The valves are classified Engineered Safety Features (Section 6.2.4.4 of Ref. 24).
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The Technical Specifications require the automatic isolation function of the valves to be tested during each refueling outage. -The leakage rate shall be tested each refueling i
outage. Because these' valves are used to inert /deinert the containment for startup and shutdown and otherwise not l
used during power operation, this is acceptable.
2 The staff recommended that Oyster Creek provide measures to ensure that isolation valve closure will not be prevented by debris which could potentially become entrained in the escaping air and steam.
The information supplied by the licensee has demonstrated that the present design using the new Clow valves presented in Section 2.2.2 meets the staff's recommendation. The drywell 4
purge inlet and exhaust penetrations are located high on the drywell walls and have radiation shields located 12 inches from their exits. The reasons, for the staff's finding that the potential for debris reaching the drywell purge / vent containment isolation valves is negligible, are the use of fast closing isolation valves (5 seconds) and the difficult paths debris must travel to reach the first isolation valve.
Debris would have to be carried up to the elevation of the penetration, travel around the radiation shields, and pass through over 20 feet of piping containing several 45 degree and 90 degree bends, before reaching the first isolation valve.
The licensee contends that the wetwell purge / vent penetrations do not require debris screens because of their location. The staff agrees with the licensee's position since the wetwell is relatively free of debris. Also, because a LOCA would occur in the drywell, the suppression pool will act to entrain any debris in the air and steam flowing from the drywell to the wetwell. The relatively small probability of debris reaching the wetwell purge / vent penetration combined with the use of fast closing isolation valves (5 seconds) constitute the basis for the staff's finding that the present measures to prevent debris from entering the wetwell purge isolation valves are acceptable.
The licensee was requested'to submit an analysis which demonstrates the acceptability of the provisions made to protect safety-related structures and equipment located beyond the purge / vent isolation valves against loss of function from the environment created by the escaping air and steam. The only safety-related equipmer.t located downstream of the purge /
vent isolation valves at Oyster Creek are the Standby Gas Treatment System (SGTS) trains. To protect the SGTS and the associated ductwork from the pressure rise that would occur before the purge valves are fully closed, the licensee by l
letter dated May 14, 1984 (Ref. 21) proposed to modify the purge / vent exhaust system by installing (1) blowout panels in the ducting (just beyond the purge system outboard isolation 2) a time l
valves and inlet valve after 5 seconds. These modifications are to be installed during the Cycle 11R outage to prevent the design pressures of the SGTS and associated ductwork from being exceeded. The staff concludes that these proposed modifications will prevent any damage from occurring to the SGTS and associated ductwork. Therefore, it is the staff's findings that the provisions to protect the safety-related structures l
l
. I downstream of the purge / vent system isolation valves are acceptable. The 5-second time delay on the o SGTS filter inlet valves has been installed (pening of the Ref. 23).
The licensee was requested to provide the amount of containment j
atmospher'a that would be released through the purge / vent l
1 solation valves before their closure in order for the staff to evaluate the radiological consequences of a LOCA occurring while purging or venting. The analyses provided by the licensee considered only the 2-inch purge / vent lines being open at the time of a LOCA, and did not include the 12,
18-and 20-inch purge / vent lines. The staff conservatively calculated the amount of containment atmosphere that would be released through all the purge / vent lines that could be open at the time of a LOCA including the large valves. This value, 2000 pound mass of steam and/or air, was used in the staff's radiological analysis of the consequences of purging at the time of a LOCA.
The staff has reviewed the radiological consequences of a i
hypothetical LOCA while purging the containment at power for Oyster Creek. This evaluation was conducted in accordance with the guidance of BTP CSB 6-4, SRP Sections 6.2.4 and 15.6.5 and Regulatory Guide 1.3.
Our evaluation is based on the bounding release estimate of 2000 i
pounds mass (lbm) of steam prior to the post-LOCA closure of the purge valves at the maximum Technical Specification primary coolant concentration of 8 uCi/gm, Dose Equivalent I-131.
It was assumed that the containment isolation would be achieved before the onset of fuel failure resulting from the accident. The 0.5%, directionally dependent, X/Q values consistent with ground level releases were used in the dose calculations. The list of apr1hable assumptions used in the analysis are given in Table i.
The staff estimated that @t ':.am release through the purge line would result in at..scretal dose of 2.8 Rem to the thyroid at the Exclusion Area Boundary (EAB) and 0.24 Rem to the thyroid at the Low Population Zone (LPZ) Boundary. The staff's Systematic Evaluation Program (SEP) Topic XV-19 dated September 1982, indicated that the LOCA doses are 3.8 Rem at the EAB and 341 Rem at the LPZ (see Table 2). The total LOCA dose including the purge dose contribution at the EAB would, therefore, meet the applicable guideline of 10 CFR Part 100.
At the LPZ, on the other hand, the negligible dose contribution from purging of 0.24 Rem would not alter the conclusion reached in the the staff's evaluation of SEP Topic XV-19 dated September 2, 1982, that because the estimated thyroid doses exceed the 10 CFR Part 100 thyroid dose LPZ guideline value by only 14% any plant backfit considerations can be pursued during the integrated assessment of Oyster Creek.
TABLE 1 ASSUMPTIONS USED TO EVALUATE THE CONTAINMENT PURGE CONTRIBUTION TO THE LOCA DOSE X/Qvalue(0-2 hour,EAB,groundlevelrelease),sec/mf7.6x10-4 (0-8 hour, LPZ, ground level release), sec/m 6.5 x 10-5 Purge valve closure time, sec less than 10 Amount of steam released through the purge valves 2000 prior to post-LOCA closure, lbm Maximum technical specification primary coolant 8
limit, dose-equivalent I-131, uCi/gm TABLE 2 RADIOLOGICAL CONSEQUENCES THYROID DOSE EAB, 0-2 HOUR LPZ, 30 DAYS Containment purge contribution 2.8 Rem 0.14 Rem
}
SEP LOCA dose 3.8 Rem 341 Rem l
Effective LOCA dose 6.6 Rem 341 Rem NOTES:
1.
The X/Q value and the LOCA dose were taken from the Systematic EvaluationProgram(SEP)TopicXV-19,datedJune 29, 1982, 2.
The whole body doses are not listed because they would be negligible when compared to the guideline values.
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The Integrated Plant Safety Assessment Report (ISPAR) for Oyster Creek, NUREG-0822, January 1983, documents the review l
completed under the staff's SEP. The IPSAR also documents the reviews to be completed under an integrated assessment of Oyster Creek.
In SEP Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment, and Topic XV-18, Radiological Consequences of a Main Steam Line Failure Outside Containment, IPSAR Sections 4.36 and 4.37, respectively, the staff presented that incorporating the BWR Standard Technical Specification (BWR-STS) limits for reactor coolant radioactivity into the Oyster Creek Technical Specifications was sufficient to ensure that the radiological consequences from these accidents were acceptably low. This would include the LOCA.
It is the staff's position in IPSAR Section 4.37 that the licensee should maintain the primary coolant radioactivity within the BWR-STS limits. The staff i
concludes that resolution of IPSAR Section 4.36 and 4.37 1
would provide the basis for acceptance of the purge dose contribution for the LPZ.
4 2.1.3 Conclusions j
Based on the above, the staff has concluded that the licensee has met BTP CSB 6-4.
Specifically, the staff has concluded that (1) the Technical Specification 3.5.A.6 which limits the use of the containment purge / vent system is sufficient control in the Technical Specifications to limit the use of this i
system, (2) the measures to prevent debris from entering the purge / vent system are acceptable, (3) the provisions to protect the safety-related structures downsteam of the purge /
vent system isolation valves are acceptable. (4) the incremental dose for the LOCA at the EAB is acceptable and 1
at LPZ will be resolved during the resolution of IPSAR Sections 4.36 and 4.37 and (5) the isolation times for the containment purge / vent valves (less than or equal to 5 seconds) are acceptable. Based on the above, the staff concludes that this component of MPA B-24 has been acceptable met and is closed.
By letter dated September 24, 1985 (Ref. 23), the licensee has proposed to cancel its modification to replace the existing large vent and purge valves in the Cycle 11R outage.
The licensee has provided the valve closure times for the existing valves in Table 6.2-12 of Reference 24. The licensee has not provided the amount of containment atmosphere that would be released through the existing purge / vent isolation valves during a LOCA before valve closure. The conservative assumption that the 20" lines are open should be made because the Technical Specifications allow these valves to be open during the run mode (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor starts up and before shutdown). The licensee should also provide justification for valves V-28-17 and V-28-18 which have valve closure times of 24.7 and 20.5 seconds which are in excess of the staff's maximum acceptable closure time of 15 seconds. This is needed by the staff for its review of the licensee's proposal.
e, t
At the resolution of this issue on the acceptability of the existing containment purge / vent valve, the licensee is requested to propose Technical Specifications on these valves listing the valve closure times because these valves may be open when the plant is in the run mode. An example of acceptable Technical Specifications is given in Enclos' re 3.
u The valve closure times should include the times listed in Table 6.2-12 of Reference 24.
2.2. Valve Operability 2.2.1 Introduction The licensee provided operability demonstration information for the purge and vent system isolation valves at its Oyster Creek Plant in References 2 and 14. The licensee's approach to operability demonstration was based on the following information presented in its submittals:
a.
References 2 and 14 stated that it is the licensee's intention to limit the opening of the containment venting and purging butterfly valves to 30* or less until such time as the valves are demonstrated to be fully qualified or replaced with fully qualified 4
valves.
b.
Other than valve identification numbers and manufacturer's names, r.one of the submittals reviewed provides analysis or test data that demonstrate operability from the fully open or 30' position.
The staff, therefore, had no basis to conclude that these valves would be operable in the event of a DBA/LOCA and requested that these valves be sealed closed except during cold shutdown or refueling.
The staff transmitted to the licensee via a 10 CFR 50.54(f) letter dated January 20, 1984, (Ref. 17) a Safety Evaluation regarding the demonstration of operability of containment purge H
and vent valves at Oyster Creek. A synopsis of that evaluation is that the licensee had submitted no information to demonstrate the operability of containment purge and vent valves. The licensee responded to the 50.54(f) letter on February 21, 1984. Firstly, the licensee enclosed its previously submitted information (August 27, 1981) on the operability of the containment purge and vent valves that the staff had not included in its evaluation. Secondly, the licensee indicated it intended to install different purge and vent valves during the Cycle 11 refueling (Cycle 11R) outage.
i The information in the August 27, 1981, submittal (Ref. 11) was not detailed sufficiently for the review being performed by the s+.aff on purge and vent valve operability. Thus, the i
I-i staff would have had to request additional details regarding the information submitted, the licensee would respond and the staff review the material, prepare a Safety Evaluation and transmit the evaluation to the licensee. To more effectively use its resources, the staff requested the detailed information required te demonstrate the operability of the new "qu'alified" containment purge and vent valves to be installed during the Cycle 11R outage. The licensee provided this information in a submittal dated April 19, 1984.
(Ref. 19)
By letter dated July 31, 1980, the licensee stated that all containment vent and purge valves would be replaced with qualified valves. The licensee requested this be delayed to Cycle 11R outage in its letter of December 24, 1981, and the staff accepted this in its letter of January 17, 1983. Until the next refueling outage the licensee could continue to use the presently installed valves provided they were blocked by mechanical means to an opening angle of 30" or less and they were used for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only during startup and prior to shutdown.
2.2.2 Requirements Demonstration of operability of the containment purge and vent valves, particularly the ability of these valves to close during a design basis accident, is necessary to assure con-tainment isolation. This demonstration of operability is required by BTP CSB 6-4 and SRP 3.10 for containment purge and vent valses which are not sealed closed during operational reactor conditions 1, 2, 3 and 4.
Description Bettis ASCO Valve Valve Size Actuator Solenoid Number (Inches)
Model Number Valve Model Number V-23-13 8
N732C-SR80 NP831664E V-23-14 8
N732C-SR80 NP831664E V-23-15 8
N732C-SR80 NP831664E V-23-16 8
N732C-SR80 NP831664E V-28-17 12 NT316B-SR2 NP831664E V-28-18 12 NT316B-SR2 NP831164E V-27-1
~18 NT420B-SR2
.NP816A74E V-27-2 18 NT4208-SR2 NP816A74E V-27-3 18 NT4208-SR2 NP816A74E V-27-4 18 NT4208-SR2 NP816A74E a
The containment purge and vent valves listed above are
)
supplied by the Clow Corporation replacing the valves originally installed at Oyster Creek. Bettis pneumatic actuators of the air to open-spring close type are provided for the valves as follows:
Valve Size Bettis Actuator (Inches)
Model Number I
8 N732C-SR80 12 NT316B-SR2 18 NT420B-SR2 l
All valves are installed outside containment.
2.2.3 Demonstration of Operability The licensee in its submittal dated April 19, 1984, provided the operability demonstration information for the containment isolation purge and vent valves at Oyster Creek.
The Clow Corporation report number 4-01-82 entitled " Purge and Vent Valve Operability Qualification Analysis" attached to the April 19, 1984 submittal contains the specifics for demonstration of valve operability.
Valve loads are predicted using model test data for the pre-diction of dynamic torque coefficients. The model tests take into account the effects of u elements.(elbows, tees, etc.)pstream and downstream piping including various separation distances and orientations. Conservative assumptions are used in the determination of dynamic torque with no credit taken for pressure ramp in containment and no credit taken for backpressure due to downstream piping.
Using model test data, dynamic torques are calculated by a computerprogramCVAP(ClowValveAnalysisProgram) developed for use in predicting valve operating characteristics.
In the computer program, mass flow rates are predicted by stan-dard equations for flow through an ideal converging nozzle adjusted with coefficients developed in the tests. Torques are predicted on the basis of the equation.
3 T=C dP D T
y where:
T = predicted aerodynamic torque (in/lb)
Cr = torque coefficient developed in model tests 2
dP = pressure differential across the valve (lb/in )
Dy = nominal valve diameter (in.)
The power of 3 used in the equation and the CVAP program is a derived value obtained by use of the equations for a general control volume. A test performed on a full size 12-inch valve indicated that torques were approximately 65% of the values obtained for the same valve from the CVAP prrgram demonstrating additional conservatism in the analysis.
Actuator spring output torques are tabulated versus valve angle in Section 2.2.3.2 of the Clow valve qualification report (April 19, 1984 submittal).
Valve dynamic torques calculated with CVAP program for each valve as installed are provided by the licensee in the April 19, 1984 submittal. A typical table is shown on the next page.
Seismic qualification for the containment purge and vent valves at Oyster Creek is demonstrated by the following seismic qualification reports prepared by Wyle Laboratories in Huntsville, Alabama.
1.
Report WR 81-53 for Clow 8-inch wafer soap valve (January 1982) Mark numbers V-23-13, -14, 16 Clow job number 80-8170-01, -01 2.
Report WR 81-54 for Clow 12-inch wafer stop valve (December 1981) Mark numbers V-28-17, -18 Clow job number 80-8170-03.
3.
Report WR 81-55 for Clow 18-inch wafer stop valve (January 1982)
Mark Numbers V-27-1, 2, 3, 4 Clow job number 80-8170-04.
Valve design and manufacturer is in accordance with:
1.
Jersey Central Power and Light Company Procurement
~
Specification number 492-7, Revision 3, dated July 29, 1981.
2.
ASME Code section III, Division 1, subsection NC and Appendices I and XIV.
2.2.4 Evaluation The April 19, 1984 submittal from the licensee which includes the Clow Corporation Report Number 4-01-82 entitled " Purge and Vent Valve Operability Qualification Analysis" contains the operability demonstration information for this evaluation.
Table 3.
Valve Number V-23-15 (18 inches).
Torque summary for valve numbers V-23-15 (18 inches).
All torques in in/lbs.
Model Test Actuator Valve Valve Torque for Torque Torque for Angle Angle Straight Flow Modification ~ ' Installed Condition
(*)
(*)
Normal Maximum Factor Normal Maximum 80 90 4
1548 2.21 9
3421 70 80 7
2798 1.56 11 4365 60 70 8
3381 1.12 9
3787 50 60 9
3472 2.05 9
3645 40 50 9
3238 1.00 9
3238 30 50 9
3238 1.00 8
2858 20 30 8
2519 1.00 8
2519 10 20 8
2405 1.00 8
2405
The Clow analysis assumes worst case postulated accident conditions; peak containment pressure taken from the LOCA containment pressure response curves with no credit taken for ramp pressure rise and single valve closure which the staff finds acceptably conservative.
Dynamic torque coefficients are derived from a model test program performed for Clow by Dr. A. L. Addy of the University of Illinois using the University's facilities.
From the data base developed by the model tests, a computer program (Clow valve analysis program - CVAP) is used to predict valve operating characteristics.
The Clow approach to dynamic torque prediction which the staff finds conservative, involves comparing the installed valve / piping configuration with the test data configuration and selecting the appropriate set of test data.
Results of the torque analysis are shown for each valve in Section 3.5.2 of the Clow analysis (Ref. 19), with torque modification factors applied to account for the effects of upstream fittings.
Table 33 from the Clow analysis (as seen in Section 2.2.3 of this safety evaluation), tabulates dynamic torque during valve closure for valve number V-23-15 (8-inch diameter).
Stress analysis results for the valve critical parts are shown in Section 4.1 of the Clow Analysis.. The analysis was 4
performed for Clow by Wyle Laboratories using the ANSYS finite element computer program. Allowable stresses are in accordance with ASME Section III requirements and Table 10 of allowed stress values. Table 11 through 17 contained in the Clow report (Ref.19) sumarize the stress analysis results for the 8, 12-and 18-inch purge and vent valves, and show that the maximum stresses calculated are less then the ASME code allowables.
The Bettis (air to open and spring to close) pneumatic actuators used with the Clow valves have ratings at the sealing position (O*) as follows:
Valve Size Actuator Sealing (Inches)
Model Number Torque at 0*
Torque i
i 8
N732C-SR80 10,060 in/lb 8,324 in/lb 12 NT316B-SR2 21,045 in/lb 20,600 in/lb i
18 NT420B-SR2 41,911 in/lb 41,000 in/lb l
The torque margin for valve sealing is adequate since inertial effects provide additional sealing torque.
i l
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Frequency analysis for the three valve sizes used at Oyster Creek shows that the lowest resonant frequencies are above
)
40 Hz.
l 8-inch valves = 122 Hz 12-inch valves = 108 Hz 18-inch valves = 124 Hz Seismic qualification for the containment purge and vent valves has been addressed by the licensee in the following reports by Wyle Laboratories. These reports are references on page 100 of the Clow analysis (April 19, 1984 submittal, Ref. 19).
1.
Report WR 81-53 for Clow 8-inch wafer stop valve (January 1982)
Mark numbers V-23-13
-14. -15, -16 Clow job number 80-8170-01, -02.
2.
Report WR 81-54 for Clow 12-inch wafer stop valve (December 1981)
Mark numbers V-28-17. -18 Clow job number 80-8170-03.
3.
Report WR 81-55 for Clow 18-inch wafer stop valve (January 1982)
Mark numbers V-27-1,
-2, -3, -4 Clow job number 80-8170-04.
2.2.5 Sumary
~
The staff has completed its review of information submitted concerning operability of the 8, 12, and 18-inch Clow containment purge and vent valves at Oyster Creek. The staff concludes that the information submitted has satisfactorily demonstrated the ability of these valves to close against the buildup of containment pressure in the event if a DBA/LOCA.
Based on the above, the staff concludes that this component of MPA B-24 has been acceptable met and'is closed.
By letter dated September 24, 1985, (Ref. 23) the licensee has proposed to cancel its modification to replace the large containment vent and purge valves by the Clow valves in the Cycle 11R outage.
This is presently under staff review to resolve before commencement of the Cycle 11R outage.
2.3 Safety Actuation Signal override This involves the review of safety actuation signal circuits to ensure that overriding of one safety actuatioii' Tgnal does s
not also cause the bypass of any other safety actuation signal.
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.On November 29, 1978, (Ref.1), the staff issued a letter,
" Containment Purging During Normal Plant Operations, to the licensee for Oyster Creek. This letter required a review of i
the design of all safety actuation signal circuits which incorporate a manual override feature to ensure the overriding of one safety actuation signal does not also cause the bypass of any other safety actuation signal. The licensee was requested to provide the results of this review and a schedule 3.
or procedure changes imposed or planned to correct any non-conforming circuits. By letters dated January 23, 1979; June 10 and July 31, 1980; and February 10 and April 9, 1981, the licensee responded to this issue (Refs. 2, 7, 8, 9 and 10). The staff and its contractor reviewed these responses and issued a Safety Evaluation (SE) to the licensee on August 5,1981, as part of the staff's resolution of SEP Topic VI-4, Containment Isolation System (Electrical).
The staff concluded in its Safety Evaluation that completion of the licensee's proposed modifications will make the design for reset and override of Engineered Safety Feature systems, including safety actuation signals, is acceptable. These modifications were the following:
(1) install radiation monitors to provide additional signals to actuate isolation of the containment purge and vent valves, (2) add a second key locked switch interlocked with the reactor mode switch for the containment nitrogen purge and vent valves so that a single failure will not disable the ability to i
purge and vent containment, (3) changeover to three-position control switches so that operator action to reset all containment isolation circuits will not be necessary, and (4) install a bypass to allow the manual trip of the core spray pumps in the presence of the low-low reactor water level trip signal.
i Based on the above, the staff concludes that this component of has been acceptably met by the licensee and is closed. Modifications (3) and (4) discussed above have been completed (Ref. 23).
Modification (1) was committed to by the licensee in its letter dated April 9, 1981 (Ref. 10) to utilize two containment high range radiation monitors to provide the isolation signal. These monitors i
will be installed in the Cycle 11R outage as required by Confirmatory Order dated June 17, 1983. Modtfication (2) has not been completed. This modification is needed for the containment nitrogen purge and vent system to meet the single failure criteria for purging the containment of combustible gases following the design basis loss-of-coolant accident. The licensee has proposed by letter dated September 24, 1985, (Ref. 23) to cancel its commitment to install a safety grade nitrogen purge and vent i
system which would thus delete this modification. This is presently under staff review to resolve before commencement of the l
Cycle 11R outage.
I The nitrogen purge and vent system is the 2-inch valves in the containment purge and vent system shown in the figure in Section 4
2.1.1 of this evaluation.
In its letter dated July 31,1980,(Ref.
- 8) the licensee comited to upgrade the nitrogen purge and vent system concurrently with the replacement of'the large containment purge and vent valves. The licensee requested that replacement of 1
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these valves be deferred to the Cycle 11R outage in its letter dated December 24, 1981, and the staff accepted this deferrment in its letter dated January 17, 1983.
2.4 Containment Leakage Due to Seal Deterioration As a result of the staff's study of valve leakage due to seal deterioration, periodic leakage integrity tests of the purge / vent isolation valves were recommended for all plants utilizing valves with resilient seals. The new purge / vent system isolation valves committed by the licensee to be installed at Oyster Creek in Cycle 11R outage utilize a metal-to-metal sealing arrangement. Therefore, i
staff recomendations related to testing of valves with resilient seals do not apply to these new valves. The leak test frequency specified in Appendix J to 10 CFR Part 50 is the only leak test frequency that would be applicable to the new Oyster Creek purge /
vent isolation valves. Based on the above, the staff concludes that this component of MPA B-24 has been acceptable met and is closed.
By letter dated September 24, 1985, the licensee has proposed to cancel its modification to replace the large vent and purge valves by the Clow valves in the Cycle 11R outage. The new Clow valves were discussed in Section 2.0 of this evaluation. The licensee stated in the enclosure to its letter that valves V-27-1, V-27-2, V-27-3 and V-27-4 have EPDM seals and valves V-28-17, V-28-18, V-23-13. V-23-14, V-23-15 and V-23-16 have nitron seals. These seals are resilient material seals. provides staff guidance on resilient seals. Upon resolution of the staff's review of the operability of the existing valves, it will be requested, if needed, that an appropriate Technical Specification change incorporating the test requirements on the existing valves with resilient seals together with the details of the test program be submitted to the staff. An acceptable Technical Specification is given in Enclosure 3.
2.5 TMI Action Plan Item II.E.4.2 Positions 1 through 4 of TMI Action Plan Item II.E.4.2, Containment Isolation Dependability, were presented and were found acceptable for Oyster Creek in the staff's Safety Evaluation dated May 8, 1980.
In that Safety Evaluation, the staff concluded that the licensee's implementation of the Category "A" Lessons Learned requirements, which included Positions 1 to 4, at Oyster Creek was acceptable.
In response to Item II.E.4.2.4, the licensee committed to replace.all two positions control switches associated with the large purge and vent valves with three position switches (Ref. 10). This modification has been completed (Ref. 23).
Position 5 Minimum Containment Pressure Setpoint, was accepted by the staff for Oyster Creek in its Safety Evaluation dated July 28, 1981.
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Position 6, Operable Valves, Or If Closed Must Be Sealed Closed, was accepted by the staff for Oyster Creek in its SE dated August 31, 1983 Ref. 16).
In this Safety Evaluation, the staff concluded Oyster Creek met Position 6 because the containment purge and vent valves met the staff Interim Position for operation in its October 23, 1979 letter (Ref. 4). This was determined in the staff's Safety Evaluation dated 2
October 8, 1980.
Position 7, Primary Containment High Radiation to Close Vent and Purge Valves, was accepted by the staff for Oyster Creek in its Safety Evaluation dated August 31, 1983 (Ref. 16). The licensee ccnnitted in its letter dated April 9,1981 (Ref.10), to utilize two containment high range radiation monitors to provide an isolation signal based l
on containment radiation to close the purge and vent valves. The deferment of implementation of Position 7 to the Cycle 11R outage was granted by Confirmatory Order dated June 17, 1983. Therefore, the staff concluded that the licensee has met the requirements for Position 7 for Oyster Creek.
Based on the above, the staff concludes that this component of MPA B-24 has been acceptable met and is closed.
By letter dated September 24, 1985 (Ref. 23) the licensee stated that its commitment to install a containment high range radiation signal to isolate the large containment vent and purge valves was being reexamined.
The licensee stated that this item would be addressed in a future submittal.
The staff has not changed its position that containment purge and vent isolation valves that may be open during an accident must close on a high radiation signal. provides staff guidance on this issue.
The justification for not installing the high radiation signal to isolate all containment vent and purge valves, large and small valves, should address the staff's concerns in Enclosure 2.
i 2.6 SEP Topic VI-4, Containment Isolation System The definition and the safety objectives of SEP Topic VI-4, Containment Isolation System, are given on pages A-50 and A-51 of the Integrated
- l Plant Safety Assessment Report (IPSAR), NUREG-0822 dated January 1983, for Oyster Creek which was issued by the staff as a part of the staff's SEP. This topic included aspects of Multi-Plant Generic Activity (MPA) B-24. The staff's final Safety Evaluations for this topic are dated August 5, 1981, and June 5, 1982.
The electrical, instrumentation and control (EIC) aspects of the override of the containment purge valve isolation was addressed in i
i the staff's Safety Evaluation dated August 5, 1981. This was the Safety Evaluation for SEP Topic VI-4, Containment Isolation System 4
(Electrical), for Oyster Creek. The staff stated that the scope of review and evaluation performed for MPA B-24 includes the electrical aspects of SEP Topic VI-4..The staff concluded that Oyster Creek met the EIC requirements of MPA B-24 and, therefore, the electrical requirements of SEP Topic VI-4. This is also discussed in Section 2.3, Safety Actuation Signal Override, above.
1
The Safety Evaluation dated August 5, 1981, on Containment Isolation System (Mechanical) did not address the mechanical aspects of MPA B-24 because the containment purge and vent valves met the staff's Interim Position for operation with these valves in its letter dated October 23, 1979. Therefore, for SEP Topic VI-4, the mechanical aspects of MPA B-24 are addressed in Sections 2.1 to 2.4, above.
Based on the above, the staff concludes that this component of MPA B-24 has been acceptable met and is closed.
3.0 CONCLUSION
1 Prior to the licensee's September 24, 1985, letter, the staff was prepared to issue its Safety Evaluation to close out MPA B-24.
With the licensee's commitments to (1) replace the containment large purge and vent valve by new Clow valves (Ref. 8), (2) to isolate the purge and vent valves by a containment high range radiation signal (Ref.10), and (3) to install a pressure relief vent in the exhaust j
duct (Ref. 21), the licensee, as presented in Sections 2.1 to 2.6
'f above, had met the requirements of MPA B-24 and the staff could close out its actions on this issue. The licensee would install the new valves, the pressure relief vent and the containment high range radiation signal to isolate the purge and vent valves in the Cycle j
11R outage.
By its letter dated September 24, 1985 (Ref. 23), the licensee' has i
proposed to cancel its comitments to install the new valves, and the pressure relief vent. These changes are presently under review by the staff to be resolved before the commencement of the Cycle 11R
]
outage. To assist the staff in its review of these issues the j
following information is needed: (1)providetheamountofcontainment i
atmosphere that may be released through the existing valves during a LOCA before valve closure and (2) provide justification for valves i
V-28-17 and V-28-18 with valve closure times greater than 15 seconds, l
the staff's actions on MPA B-24. The staff's review of the licensee's i;
September 24, 1985, letter will be separate plant-specific actions.
!j The separate plant-specific actions will include the staff's review 1
of the licensee's request in the September 24, 1985, letter to cancel the commitment to install a containment nitrogen purge and i
j vent system to comply with 10 CFR Part 50.44(g). This review sill consider the modification to add a second keylocked switch for this i
system to assure that a single failure will not disable the ability to purge and vent containment. This modification is discussed in i
Section 2.3 of the enclosed evaluation.
}
Enclosures 1 to 3 which are attached to this Safety Evaluation on MPA B-24 provide the following guidance: (1) amplification of Branch Technical Position CSB 6-4 on testin of resilient seals for the containment purge and vent valves, (g) high radiation isolation signals 2
for containment purge and vent valves and (3) acceptable Technical Specifications for valves with resilient seals and listing valve closure times. The resolution of IPSAR Sections 4.36 and 4.37 will provide the resolution of the LOCA dose at the LPZ.
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4
4.0 REFERENCES
FOR MPA B-24 1.
Letter, Dennis Ziemann (NRC) to I. R. Finfrock, Jr., (JCP&L) dated November 29, 1978; 2.
Letter, I. R. Finfrock, Jr. to Director of Nuclear Reactor Regulation, dated January 23, 1979; 3.
Letter, Darrell G. Eisenhut (NRC) to All Light Water Reactors, dated September 27, 1979; 4.
Letter, Dennis Ziemann to I. R. Finfrock, Jr., dated October 23, 1979; 5.
Letter, I. R. Finfrock, Jr. to Director, Nuclear Reactor Regulation, dated December 17, 1979; i
6.
i.etter, Dennis Crutchfield to I. R. Finfrock, Jr. dated May 8,1980.
1 7.
Letter, I. R. Finfrock, Jr. to Darrell G. Eisenhut, dated June 10, 1980; 8.
Letter, I. R. Finfrock, Jr. to Director, Nuclear Reactor Regulation, dated July 31, 1980; 9.
Letter, I. R. Finfrock, Jr. to D. G. Eisenhut, NUREG-0737, February 10, 1981;
- 10. Letter,1. R. Finfrock, Jr. to Dennis M. Crutchfield, dated April 9, 1981; i
- 11. Letter, Philip R. Clark (GPUN) to Director, Nuclear Reactor Regulation, dated August 27, 1981;
- 12. Letter, D. M. Crutchfield to P. B. Fiedler, dated January 4,1982;
- 13. Letter, P. B. Fiedler to D. M. Crutchfield dated July 15, 1982;
- 14. Letter, P. B. Fiedler to D. M. Crutchfield, dated January 13, 1983;
- 15. Letter, P. B. Fiedler to Darrell G. Eisenhut, dated April 15, 1983;
- 16. Letter, D. M. Crutchfield to P. B. Fiedler dated August 31, 1983;
- 17. Letter, D. M. Crutchfield to P. B. Fiedler, dated January 20, 1984;
- 18. Letter, P. B. Fiedler to D. M. Crutchfield, dated February 21, 1984;
- 19. Letter, P. B. Fiedler to D. M. Crutchfield, dated April 19, 1984;
- 20. NEDO 22155, " Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containments", dated June 1982; I
~
- 21. Letter, P. B. Fiedler to D. M. Crutchfield, dated May 14, 1984;
- 22. Letter, P. B. Fiedler to Director, Nuclear Reactor Regulation, dated July 19, 1984.
- 23. Letter, P. B. Fiedler to John A. Zwolinski, dated September 24, 1985.
- 24. Oyster Creek Nuclear Generating Station, Updated Final Safety Analysis Report, December 1984.
Principal Contributors: K..Dempsey, M. Fields, R. Scholl, R. Wright and J. Donohew.
Dated: January 21, 1986.
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9 PURGE / VENT VALVE LEAKAGE TESTS The long tem resolution of Generic Issue B-24, " Containment Purging During Normal Plant Operation," includes, in part, the implementation of that provisions should be made for leakage rate testing of the (pecifie Item B.4 of Branch Technical Position (BTP) CSB 6-4.
Item B.4 s purge / vent system) isolation valves, individually, during reactor operation. Although Item B.4 does not address the testing frequency, Appendix J to 10 CFR Part 50 specifies a maximum test interval of 2 years.
As a result of the numerous reports on unsatisfactory performance of the
~
resilient seats for the isolation valves in containment purge and vent lines (addressed in DIE Circular 77-11, dated September 6,1977), Generic Issue B-20. " Containment Leakage Due to Seal Deterioration," was established to evaluate the matter and establish an appropriate testing frequency for the isolation valves. Excessive leakage past the resilient seats of isolation valves in purge / vent lines is typically caused by severe environmental con-l ditions and/or wear due to frequent use. Consequently, the leakage test frequency for these valves should be keyed to the occurrence of severe environ-mental conditions and the use of the valves, rather than the current require-ments of 10 CFR 50, Appendix J.
It is recommended that the following provision be added to the Technical Specifications for the leak testing of purge / vent line isolation valves:
" Leakage integrity tests shall be performed on the containment isolation valves with resilient material seals in (a) active purge / vent systems (i.e., those which may be operated during months and (b)g Modes 1 through 4) at least once every threepassive plant operatin f
administrative 1y controlled closed during reactor operating Modes 1 through 4) at least once every six months."
By way of clarification, the above proposed surveillance specification is predicated on our expectation that a plant would have a need to 'go to cold shutdown several times a year. To cover the possibility that this may not occur, a maximum test interval of 6 months is specified. However, it is not our intent to require a plant to shutdown just to conduct the valve If licensees anticipate long duration power oper-leakage integrity tests.
ations with infrequent shutdown, then installation of a leak test connection that is accessible from outside containment may be app opriate. This e
It will not be will pemit simultaneous testing of the redundant valves.
possible to satisfy explicitly the guidance of Item B.4 of BTP CSB 6-4 (which states that valves should be tested individually), but at least some testing of the valves during reactor operation will be possible.
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It is intended that the above proposed surveillance specification be applied t
i to the active purge / vent lines, as well as passive purge lines:
1.e., the purge lines that are administrative 1y controlled closed during reactor oper-ating modes 1-4.
The reason for including the passive purge lines is that B-20 is concerned wtih the potential adverse effect of seasonal weather con-ditions on the integrity of the isolation valves. Consequently, passive purge lines must also be included in the surveillance program.
The purpose of the leakage integrity tests of the isolation valves in the containment purge and vent lines is to identify excessive degradation of the resilient seats for these valves. Therefore, they need not be conducted with the precision required for the Type C isolation valve tests in 10 CFR These tests would be performed in addition to the Part 50, Appendix J.
I quantitative Type C tests required by Appendix J and would not relieve the licensee of the responsibility to conform to the requirements of Appendix J.
j In view of the wide variety of valve types and seating materials, the acceptance criteria for such tests should be developed on a plant-specific
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CONTAINMENT SYSTEMS BRANCH ENCLOSURE 2 UPDATED EVALUATION OF BWR OWNERS GROUP POSITION ON ITEM II.E.4.2.(7) 0F NUREG-0737 i
In NUREG-0737. Item II.E.4.2.(7),'we state that containment purge and vent isolation valves must close on a high radiation signal.
The BWR Owners Group has perfomed an assessment to determine the benefits of providing automatic closure of the containment vent and purge valves on a containment high radiation signal. This assessment, contained in a letter from T. J. Dente i
J j
to D. G. Eisenhut, dated June 29, 1981, concludes that this automatic closure on a high radiation signal will not appreciably alter the probability for significant releases of radioactivity through these lines. 'The bases for the 1
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BWR Owners Group conclusion relies on th'e following points:
1 1.
Automatic isolation is already achieved through diverse inputs i
(high dry-well pressure and low' reactor water level);
2.
Thecontainmentventandpurgevalvesarenormailyclosed; 3.
Several diverse methods exist for detection of primary coolant boundary leakage that could indicate to the operator that a high radiation condition in the containment may exist; and 4.
Pipe breaks leading to leakage rates less than the Technical Specification limits that are not innediately isolated by the operator result in offsite doses less than 10 CFR Part 100 dose limits.
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i It is the staff's position that the above arguments for not having a high radiation isolation signal for the containmen' vent and purge valves are t
inadequate. The staff strongly believes that these valves should be isolated on the bases of a direct measurement of the parameter that the containment isolation system is designed to protect the public from, i.e., radiation.
This view is based on the potentially greater impact on offsite doses relative to releases through other lines penetrating the containment, since the vent i
and purge lines provide a direct path from the containment atmosphere to the j
environs. The staff's view is that having only indirect parameters as isolation i
signals, such as high drywell pressure or low reactor water level, is insufficient for assuring that these valves will close in a timely manner.
The argument that containment vent and purge valves are normally closed and, l
therefore, do not reqyb a high radiation isolation signal is insufficient because these valves are normally open during startup and shutdown. Since l
these are transient conditions, we would expect at least as high a likelihood i
of a release occurring during these periods as during steady state periods.
l Moreover, since it is essential for the containment vent and purge valves to l
receive timely isolation signals under these circumstances, the staff's i
position is that a high, radiation isolation signal'is needed to accomplish this function.
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r Reliance on operdtor actioh to close the containment vent and purge valves r
is not acceptable because of the delays that could occur while the*cperator l
1s handling matters more directly related to the initiating event.
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In response to the argument that leakages less than the Technical Specification limits produce low offsite doses, the staff feels that the purpose of adding a high radiation isolation signal to the containment vent and purge valves is
~
to protect against substantial releases of radiation (10 CFR Part 100 dose limits) for accident conditions while for nonnal conditions (e.g., leakages less than Technical Specification limits), the purpose of these valves is to close before 10 CFR Part 20 dose limits are exceeded.
In sunnary, it is the staff's position that all containment vent and purge valves in lines that are used during startup, normal operation, and shutdown of the plant be provided with a high radiation isolation signal. The range and sensitivity of the radiation monitors used for this purpose shall be sufficient to assure timely closure of the vent and purge valves under both accidentconditions(limitingoffsitedosestolessthan10CFRPart100 4
guidelines) and normal operating conditions (limiting offsite doses to less than 10 CFR Part 20 limits). The high radiation signal may be either safety i
grade equipment or non-safety grade equipment. Our aim is to have a high radiation isolation sigr.a1 operable at the earliest possible time. Technical Specifications are needed consistent with the staff guidance provided in Generic Letter 82-16 dated September 20, 1982.
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NUREG-0737. ITEM II.E.4.2(7)
HIGH RADIATION ISOLATION SIGNAL FOR SMALL PURGE AND' VENT VALVES
.l SUM ARY OF KEY EVENTS l'
s The requirement' to isolate' purge / vent systems that are ope'n during normal power operation on a high ' radiation sianal, was first s'pecified in NUpEG-0737 for plants with operating licenses.
ItemII.E.4.2(7)of NUREG-0737, published in November 1980, states:
1
" Containment purge'and vent isolation valves must close on a high i
radiation signal."
In a letter dated June 29,1981, fromT.Dente,BWROwnersGroup(RWROG),
i to D. Eisenhut, NRC, the BWR06 provided an evaluation of the need for a
.high radiation isolation signal for the plants they represented. Based i
.on thir evaluation, the BWR06 concluded that such a signal was not i
.necessary. The staff reviewed this infomation and detemined that the BWR0G rationale for not installing a radiation isolation signal on purge i
and vent valves was not acceptable. This findino by the staff was transmitted to the RWROG in a letter dated October 14, 1981, from D.
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Eisenhut NRC, to T. Dente, BWROG.
I The Octobe-14, 1981 NRC letter explicitly exempted small (less than or equal to 3 ich lines) purge and vent valves from meeting the requirement j
of II.E.4.2(7), which was an error. This error was identified and...
corrected during a meeting between the NRC and representatives of BWROG, f
which was held on November 19, 1981, to discuss the results of the staff's review. During that meeting, the staff pointed out this error and stated that the earlier ESF Electrical Override Safety Evaluations were correct in '
that all open valves in lines, regardless of size, which might communicate 4
direcTTy between the containment atmosphere and the outside environment, should automatically shut on a high radiation signal. A summary of this meeting was issued on December 3, 1981, by E. Reeves. NPC, and was distributed to the BWROG. A copy of the meeting summary is provided as In late 1981 and early 1982, a modified version of the s'taffs evaluation -
contained in the October 14, 1981 NRC letter tn the BWR06, was individually transmitted to many of the' utilities comprising the BWR06. The modified
}
version removed 'the statements that explicitly exempted small purce and vent isolation valves from the high radiation isolation requirement. At 1
that time, the staff believed that this position was fully understood by i
the BWROG.
i i
In early 1984, the staff discovered that at least one BWROG utility, was-j interpreting the modified version of the staff's original evaluation as exempting small lines from the requirement of II.E.4.2f7). Closer -
l l
. examination of the modified version revealed several places.in the evaluation'where someone with no knowledge of the past history on this issue could infer that small lines still need not have a hioh radiation closure signal. Although the staff felt that the previous discussion and documentation of this issue had clarified the staff's position adequately.
another revision of the original October 14, 1981, was prepared. This revision was sent during 1984 to those PWR0G members which the staff believed may be in noncompliance with the requirements for small purge and went lines.
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CONTAINMENT SYSTEMS O
l LIMITING CONDITION FOR OPERATION I
3.6.1.7 The containment purge supply and exhaust isolation valves may be open for safety-related reasons or shall be closed. The containment
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vent line isolation valves may be open for safety-related reasons or shall be closed.
r APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
(For plants with valves closed by technical specification)
With one containment purge supply and/or one exhaust isolation valve open, close the open valve (s) within one hour or be in at least HOT STANDY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following
)
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
(For plants with valves that may be opened by technical specifications)
With one containment purge supply and/or one exhaust isolation or vent 1.
valve inoperable, close the associated OPERABLE valve and either restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or lock the p
OPERABLE valve closed.
Operation may then continue until perfomance of the next required 2.
valve test provided that the OPERABLE valve is verified to be locked closed at least once per 31 days.
Othemise, be in at least HOT STANDBY within the next six hours and 3.
in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The provisions of Specification 3.0.4 are not applicable.
4.
SURVEILLANCE REQUIREMENTS i
-inch containment purge supply and exhaust isolation valves 4.6.1.7.1 The
-incli~ vent line isolation valves shall be detemined closed at and the least onTe per 31 days.
4.6.1.7.2 The valve seals of the purge supply and exhaust isolation valves and the vent line isolation valves shall be replaced at least one per,,,, years.
3/4 6-10 1
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CONTAINMENT SYSTEMS 3/4 4.6.3 CONTAINMENT ISOLATION VALVES
.i LIMITING CONDITION FOR OPERATION j J i
1 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be I
OPERABLE with isolation times as shown in Table 3.6-1.
l APPLICABILITY: MDDES 1, 2, 3 and 4.
i ACTION:
With one or more of the isolation valves (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration I
that is open and either:
a.
Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or i
b.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least c.
f one closed manual valve or blind flange; or d.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
p SURVEILLANCE REQUIREENTS l
i 4.6.3.1 The' isolation valves specified in Table 3.6-1 shall be demonstated b
OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control I
or power circuit by performance of a cycling test, and verification of isola-tion time i
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i CONTAINMENT SYSTEMS g
i SURVEILLANCE REQUIREMENTS (Continued) i 4.6.3.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated F
OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 I
months by:
a.
Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
I b.
Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
L 4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
4.6.3.4 The containment purge and vent isolation valves shall be demonstated OPERABLE at intervals not to exceed months. Valve OPERABILITY shall be determined by verifying that when the~measuued leakage rate is added to the leakage I
rates determined pursuant to Specification 4.6.1.2.d for all other Type B and 4
C penetration, the combined leakage rate is less than or equal to 0.60La.
However, the leakage rate for the containment purge and vent isolation valves shall be compared to the previously measured leakage rate to detect excessive valve degradation.
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