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UNITED STATES [ , 7, NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 g -{ S / December 31, 1991 y.....f MEMORANDUM FOR: Paul Boehnert, Senior Staff Engineer Advisory Committee on Reactor Safeguards FROM: Christopher I. Grimes, Chief Technical Specifications Branch Division of Operational Events Assessment NRR
SUBJECT:
MONITORING AND LIMITATIONS FOR RCS LEAKAGE OUTSIDE CONTAINMENT During the STS presentation to ACRS on Friday, December 13, 1991. Dr. Michelson requested information on the monitoring and limitations for RCS leakage from systems outside containment for BWRs. A review of the current BWR/4 and BWR/6 STS shows two specifications which address such leakage monitoring and limits. LCO 3.4.8 (enclosure 1) maintains the structural integrity of all ASME Code Class 1, 2 and 3 components. This is done through the performance of the 1, 2 and 3 components. This is done through the performance of the surveillance requirements for inservice inspection and testing required by Section XI of the ASME Boiler and Pressure Vessel Code (SR 4.0.5, Enclosure 21 This specification will be a program in the Administrative Controls Section of the new STS.. Specification 6.8.4.a (Enclosure 3) in the Administrative Controls Section of the current STS establishes a program to reduce leakage from systems outside of containment which could contain highly radioactive fluids during serious transients or accidents. This specification implements ites III.D.1.1 of NUREG-7037. The progran requires preventive maintenance, periodic visual inspection and leak testing at each refueling outage. NUREG-0737 gives examples of systems which should be included in the program such as residual heat removal, containment spray recirculation and high pressure injection recirculation. The emphasis appears to be on systems which would be moving highly contaminated fluids around outside the containment under accident conditions. This program will be retained in the Administrative Controls Section of the new STS. We examined the TS of 23 BWRs to see if the Reactor Water Cleanup Systes ! (RWCU) is included in the program. Most of the plants do have the program; I I however, some (5 of 23) of the old, custom TS did not have the program in the j administrative controls part of the TS. Some plants specified that the RWCC was included in the program (5 of 23). Some plants did not list the specific systems included in the program (4 of 23). Finally, some plants did list the systems covered by the program, and RWCU was not on the list (9 of 23). s 9807100067 980624 PDR FOIA
,UNNERST98-155 PDR _
e . . Enclosure 1 REACTOR COOLANT SYSTEM e 3/4.4.8 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.8. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5. ACTION:
- a. With the structural integrity of any ASME Code Class I component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NOT considerations.
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant
. System temperature above 200*F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
! SURVEILLANCE REQUIREMENTS i 4.4.8 No requirements other than Specification 4.0.5. i i GE-STS (BWR/6) 3/4 4-23 ,
Enclosure 2 APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIOEL CONDITIONS or other conditions specified for individual Limiting Conditjens for Operation unless otherwise stated in an individual Surve111anct Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 255 of the surveillance
- interval, but
- b. The combir.ed time interval for any 3 consecutive survefl hrce intervals shall not exceed 3.25 times the specified survatillance interval.
4.0.3 Failure to perfo m a Surveillance Requirement within the allored surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a limiting Condition for.0peration. The time limits of the ACTION requirements are applicable at the time it is identified that a surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours to pemit the completion of the surveillance when the allowable outane time limits of the ACTION requirements are less than 24_ hours. 5urve1' lance Requirements do not have to be performed on Inopefable equipment.. 4.0.4 Entry into an OPERATIONAL CONDITION or other specified ap condition shall not be made unless the Surveillance Requirement (plicable s) associated with the Limiting Condition for Operation have been performed within the
- applicable surveillance interval or as otherwise specified. This provision , shall not >revent passage through or to 0PERATIONAL CONDITION 5 as required tn comply witi ACTION requirements.
I 4.0.5 Surveillance Requirements for inservice inspection and testing of ASPE Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves -
l shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and l BWR STs 3/4.0-2
a . . APPLICABILITY SURVEILLANCE REQUIREMENTS Pressure Yessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications: l ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At.laast once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days
- c. The. provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be-construed to supersede the requirements pf any Technical Specification.
BWR STS 3/4.0-3 %
s . . Enclosure 3 ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS t 6.8.1 ' Written procedures shall be established, implemented, and maintainec covering the activities referenced below:
- a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2 February 1978.
- b. The applicable procedures required to implement the requirements of NUREG-0737.
- c. Refueling operations.
- d. Surveillance and test activities of safety-related equipment.
- e. Security Plan implementation.
- f. Emergency Plan implementation.
- g. Fire Protection Program implementation.
6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed by the (URG) and shall be approved by the (Plant Superintendent) prior to implementation and reviewed periodically as set forth in administrative pro-cedures. 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made pro-vided: ,
- a. The intent of the original procedure is not altered;
- b. The change is approved by two members of the unit management staff, at least one of whom holds a Senior Operator license on the unit affected; and
- c. The change is documented, reviewed by the (URG), and approved by the (Plant Superintendent) within 14 days of implementation.
6.8.4 The following programs shall be established,' implemented, and maintained:
- a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside l
containment that could contain highly radioactive fluids during a , serious transient or accident to as low as practical levels. The i systems include the (HPCI, CS, RHR, RCIC, hydrogen recombiner, l process sampling, containment and standby gas treatment) systems. The program shall include the following:
- 1. Preventive maintenance and periodic visual inspection requirements, and .
I
- 2. Integrated leak test requirements for each system at refueling cycle intervals or less.
GE-STS 6-13 9 k.
* . . T rim Le e w s w .i - ... e y IESES FOR LIMITING CDt TIONS FOR OPERATICN AW SURVEL JG RICUIRDENTS I
3.6.K STRUC'IGAL INTERITY In-service inspection of ASME Code Class 1, 2, and 3 (ensivalent) conponents and in-service testing of ASME Code Class 1, 2, and 3 (ocuivalent) panps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressare Vessel Cbde and Addenda as reo.lired bf 10CTR50.55a(g) . 'Ihis objective will maintain the structural integrity of safety-related caponents, pinps, and valves which are necessary to safely shat down the plant or mitigate the cmseatences of an accident. Amendment NO.119 3.6-23 g "I
q l' . ADMINISTRATIVE CONTROL 1 to enter such areas shall be provided with or accompanied by one or more of the following.
- 4. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the i
radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them. ! c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Feelth l Physics supervision in the Radiation Work Permit. 6.12.2 The reovirements of 6.12.1, above, shall also apply to each high , l' i radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition. locked doors shall be provided to prevent unauthorized i ertry into such areas and the keys shall be maintained under the administrative l control of the Shift Supervisor on duty and/or the Laboratory Foreman on duty. j j 6.13 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT . The licensee shall implement a program to reduce leakage from systems outside l containment that vould or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program ' shall include the following:
- 1) Provisions establishing preventive maintenance and periodic visual inspection requirements, and
- 2) system leakage test requirements, to the extent permitted by syrtes design and radiological conditions, for each system at a frequency not to exceed refueling cycle intervals. The systems subject to this testing are (1) Residual Heat Removal. (2) Core Spray, (3) Reactor Water Cleanup.
(4) HPCI, and (5) RCIC. 6.14 10 DINE MONITORING The licensee shall implement a program which will ensure the capability to accurately deternine the airborne iodine concentration in vital areas
- under accident conditions. This program shall include the following:
1p Training of personnel. 2 Procedures for monitoring, and 3 )f Provisions for maintenance of sampling and analysis equipment.
- Areas requiring personnel access for estabitshing hot shutdown condition.
6 21 Amenenent Ilo. 85, 79,109 HATCH - UNIT 1 6.C#ECetL4/) E ZZ[ g 3 cp B$
I i REACTOR COOLANT SYSTEM 3/4.4.8 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1. 2 and 3 components shall be maintained in accordance with Specification 4.4.8. APPLICABILITY: CONDITIONS 1. 2, 3, 4 and 5. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component not confoming to the above requirements. restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor i Coolant System temperature more than 50'F above the minimum .
temperature required by NDT considerations. .
- b. With the structural integrity of any ASME Code Class 2 com- !
ponent(s) not confoming to the above requirements, restore e the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 212'F.
- c. With the structural integrity'of any ASME Code Class 3 com-ponent(s) not confoming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
- e. The provisions of Specification 3.0.3 'are not applicable in
,. OPERATIONAL CONDITION 5. l SURVEILLANCE REQUIREMENTS 4.4.8 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated per the requirements of Specification 4.0.5.
ATCH - UNIT 2 3/4 4-20 g e ,
~ 1 l I AMIN 15tpATIVE CONTROL !
'm.
ll to enter such areas shall be provided with or accompanied by one or more of the following. j
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area,
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring
. device may be made after the dose rate level in the area has been i established and personnel have been made knowledgeable of them. '
- c. An individual qualified in radiation protection procedures who is eouipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation ,
surveillance at the frecuency specified by the facility Health l ; Physics supervision in the Radiation Work Perwit. 1 6.12.2 The requirements of 6.12.1, above, shall also apply to each high - i radiation area in which the intensity of radiation is greater than 1000 i erem/hr. In addition locked doors shall be provided to prevent unauthorized ; ertry into such areas and the keys shall be maintained under the seninistrative 1 control of the Shift Supervisor on duty and/or the Laboratory Foreman on cuty. j l 6.13 1NTEGRTTY OF SYSTEMS OUTSIDE CONTAINMENT The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a i serious trarsient or accident to as low as practical levels. This program . shall include the following:
- 1) Provisie'ns establishing preventive maintenance and periodic visual inspection requirements, and
- 2) System leakage test requirements, to the extent permitted by system design and radiological conditions, for each system at a frequency not to exceed refueling c,vele intervals. The systems sub. ject to this testing are (1) Residual Heat Removal. (2) Core Spray, (3) Reactor Water Cleanup.
- (4)HPCI,and(5)RCIC. .
e e l - UNIT 6-19 Amendment 100. M e 47 . C.ccoor@ 11+nL ef 8 ep 45
+ . .
J. M
-2 December 31, 1991 During the ACRS presentation Dr. Michelson specifically asked about Hatch plant; the Hatch specifications in the two areas addressed above are enclosed (Enclosure 4).
Richard L. Emch, Jr. for Christopher I. Grimes, Chief Technical Specifications Branch Division of Operational Events Assessment, NRR
Enclosures:
As stated DISTRIBUTION: WTRussell OTSB R/F DOEA R/F Central Eiles CERossi RLEsch CIGrimes CHBerlinger AEChaffee FMReinhart CLHoxie CMAbbat e RJGiardina , DOCUMENT NAME: C:LEAKRCS.RJG PMC's PC OTSB:DOEA:NRR C:0TSB:D0 : 'RR M JGlardi g OTSB:DOEAg[RR RLEnch N IGrime [ V 12/jy/91 12/3//91 2/J//91 s
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