ML20199J459
| ML20199J459 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 01/16/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199J454 | List: |
| References | |
| NUDOCS 9802050317 | |
| Download: ML20199J459 (5) | |
Text
-.
@t%
p l'a UNITED STATES
}
NUCLEAR REGULATORY COMMISSION s*
2 WASHINGTON, D.C. - aaat SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.12KO FACILITY OPERATING LICENSE NO. NPF 57 PJ&ADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION. UNIT 1 DOCKET NO. 50 352
1.0 INTRODUCTION
By letter dated October 24,1997. Philadelphia Electric Company (the licensee) submitted a request for changes to the Limerick Generating Station (LGS), Unit 1, Technical Specifications (TSs). The requested changes would permit operation of the facility with control rod 50-27 uncoupled for the remainder of Cycle 7 for LGS Unit 1. Specifically, the change would revise Section 3/4.1.3.6 of the LGS Unit 1 TS to exempt control rod 50-27 from the coupling test for the remainder of Cycle 7, provided certain con @ns are met. Also, repositioning of the control rod to its proposed step 46 position wo.ald be allowed when the reactor power is above 10% of rated power. The surveillance requirements of TS 4.1.3.6 d would include additional monitoring during the repositioning evolution. In this regard, neutron monitoring by means of either the Local Power Range Monitor (LPRM) or the Traversing incore Probe (TIP) systems would be used to verify the control rod movement.
On August 31,1997, during the weekly exercise test of control blade 50-27, when the blade was inserted one notch from the full-out position 48 to position 46 and then retumed to full out, the full out position indication of 48 was lost. A continuous withdrawal signal was selected to restore position indication which resulted in a " rod overtravel" alarm, indicating decoupling of the drive from the control blade, in accordance with existing procedures, power was reduced to below 60%, followed by fullinsertion of blade 50 27 to position 00 and its being disarmed. Upon subsequent troubleshooting, the blade was fully withdrawn (again, without receiving position indication for the full-out position of 48). Blade movement was verified during the withdrawal by monitoring LPRMs. The blade was judged to be coupled since an overtravel alarm was not received following selection of a continuous withdrawal signal after the blade was judged to be fully withdrawn; however, since the full-out position indication of 48 was not received, the blade could not procedurally be determined to be coupled and was therefore re inserted fully to position 00 and disabled by hydraulic block in accordance with TS requirements.
Limiting Condition for Operation 3.1.3.6.a1 requires that if recoupling is not accomplished in the first attempt, the rod is to be declared inoperable, fully inserted, and disabtr,d electrically or hydraulically. Surveillance Requirement 4.1.3.6.b. requires that the coupling test be performed any time the control rod is withdrawn to the full out position. The only way to verify rod coupling,
~
9802050317 900116 PDR ADOCK 05000352 P
2 as described in existing plant procedures (Control Rod Exercise and Control Rod Coupling Check) is to fully withenew the blade to position 48, and then apply a continuous withdrawal signal to vertfy that no overtravel alarm occurs. By the nature of the problem with blade 60 27, an indication of 48 cannot be received; hence, it cannot be positively ascertained that the blade is fully out in order to apply the withdrawal signal to check for overtravel. Since the blade can not be assured of being coupled, it must be assumed to be not coupled. The proposed TS change would suspend the requirement to fully insert and disarm the subject blade for the remainder of Cycle 7 at LOS Unit 1.
2.0 PROPOSED T8 CHANGES 2.1 Limiting Condition for Operation 3.1.3.6.a.2. has the underlined phrase added:
If recoupling is not accomplished on the first attempt or, if not pom16tted by the RWM, then except as in 3.1.3.6.d or until permitted by the RWM, declare the control rod Inopersole, insert the control rod and disarm the associated directional control valves ** either:
(The balance of this paragraph remains unchangea.)
2.2 Limiting Condition for Operation action 3.1.3.6.d. is added; d.
For control rod 50-27, for the remainder of Unit 1 Cycle 7. If coupling can not be established the uncoupled rod may be withdrawn when rated thermal power exceeds 10% only if all the following conditions are satisfied:
1)
TN uncoupled control rod may not be withdrawn ps,st notch pusition 46, and 2)
No other uncoupled control rod is withdrawn.
2.3
- Surveillance Requirement 4.1.3.6.d. is added:
d.
When repositioning the uncouplad control rod per Specification 3.1.3.65 the uncoupled rod's position shall be venfied to have followed the control rod drive by neutron instrumentation (LPRM or TIP). If the control blade can not be verified to have followed the drive out to its final position, then the rod shall be completely inserted and the control rod directional valves disarmed as stated in3.1.3.6.a.2.
- 0 EVALUATION The licensee's attempt to verify that the coupling integrity of rod 50 27 was unsuccessful, and, therefore, it must be assumed that the control rod and drive are uncoupled. The primary concem for control rod coupling integrity is its impact on the potential increase in the probability of a
=
N
3 control rod drop accident (CRDA) as analyzed in the Final Safety Analysis Report. Additionally, control rod coupling integrity ensures that the indicated control rod position is indicative of the
- actual control rod position. The uncoupled rod condition also raises an operational concem for equipment damage due to scram loading. The rod could separate from the control rod crive (CRD) during the deceleration phase of the scram stroke, which could resuN in increased loads on the affected parts.
The licensee has proposed procedural changes to assure that withdrawal operation with the i
uncoupled rod 50 27 will not pose a significant CRDA conoom for the remainder of the fuel Cycle
- 7. The rod will remain inserted and not be withdrawn at reactor power levels less than 10
- percent. During the withdrawal sequence above 10 percent, neutron flux information in the vicinity of the rod will be monitored to verify that the control rod blade tracks with the drive movement. This will ensure that the rod is not sticking and separated from the CRD.
An analysis performed by General Electric Company (GE) for the licensee shows that, for the fuel cycle under consideration, the consequences of a CRDA at power levels above 10 percent of rat 6d thermal power are negligible and that no constraints on control rod sequences are required.
- Below 10 percent of rated power, the uncoupled control rod will be fully inserted. Above 10 percent power, the compensatory actions ensure that the Rod Block Monitor mitigates the consequences of a Rod Withdrawal error.
The licensee's proposed chsnges provide an additional measure to minimize the possibility of a CRDA by requiring the use of neutron instrumentation (LRPM or TIP) to verify rod position during repositioning of the uncoupled rod. This is addressed in the prop.ted change to Surveillance Requirement 4.1.3.6.d.
The GE analysis also addressed the possibility of equipment damage from scram loadings.
Mechanism damage could occur during the deceleration phase of the scram stroke. The uncoupled rod would continue to move upward and tl's velocity limiter would strike the bottom of the fuel support casting. Analysis shows that in this scenario, damage might occur to the velocity.
limier, or upon rebound, to the spud and the lock plug. However, there is insufficient energy to dislodge the fuel support and fuel, i
GE has provided recommended operating strategies to minimize poss'ble scram load problems.
The recommended operation with rod 50 27 withdrawal limited to notch position 46 minimizes the :
scram loadings on the spud and socket. The weekly tests of rod movement required by the TSs will continue, thus assuring rod movement capability.
w I
j
=2.-.-
4 GE has evaluated the effect of the control rod coupling integrity on scram performance. The scram and insertion performance were not considered to be degraded nor would other reactivity control functions be adversely affected. Since the rod will be operated at a slightly inserted position for full withdrawal, it should have slighth better scram reactivity insertion characteristics.
It is therefore reasonable to conclude that opeistion with rod 50 27 fuly withdrawn will not lead to any condition adverse to reactor safety.
The TS changes socompanying this mode of operation consist of changes to Sections 3.1.3.6 and 4.1.3.6. With the change to Section 3.1.3.6, rod 50 27 may be withdrawn when rated thermal power is gmater or equal to 10 percent under certain conditions. These conditions are that no other uncoupled rods are withdrawn rand rod 50 27 may not be withdrawn past notch position 46. The change to Section 4.1.3.6 requires the use of neutron instrumentation (LPRM or TIP) to verify that rod 50 27 followed the CDR during mpositioning. If the blade cannot be verified to have followed the drive to its final position, then the rod shall be completeh inserted and the control rod directional valves disarmed, as stated in Section 3.1.3.6.a.2. These TS changes adequately implement the required changes in rod operation and are acceptable.
4.0
SUMMARY
CONCLUSIONS Based on the NRC staff review of the licensee's submittalin support of the proposed TS changes in the operation of control rod 50 27 for the remainder of Cycle 7 for LGS Unit 1, we find the proposed amendment acceptable.
5.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with rupct to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes the surveillance requ rements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in toe types, of any effluents that may be released offsjte, and that there is no significant increase in individual or cumulative occupational radiation exposum. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 61644). Accordingh, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(g). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment nood be prepared in connection with the isstance of the amendment.
s 5
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed ti.anner, (2) such activities will be conducted in compliance with the Commi,sion's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: H. Richings B. Buckley Date:
January 16, 1998
,