ML20147B540

Forwards Response to NRC 871222 Request for Addl Info Re Util License Change Application 159 Concerning Proposed Changes to Tech Spec 3.9.9, Containment Ventilation Isolation Sys

ML20147B540
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	02/25/1988
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-66638, NUDOCS 8803020115
Download: ML20147B540 (8)

Text

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M s Pbrtland General ElectricCoirpiny David W. Cockfield Vice President, Nuclear February 25, 1988 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555

Dear Sir:

License Change Application 159

Reference:

NRC to PGE Letter License Change Application 159 (TAC 66638),

December 22, 1987 Pursuant to the above reference, responses to questions on License Change Application (LCA) 159 are provided in Attachment 1. This LCA proposed changes to Trojan Technical Specification 3.9.9 (Containment Ventilation Isolation System) to resolve inconsistencies within the Technical Specifi-cations. The responses to the questions have been incorporated into the significant hazards consideration determination for the LCA as appropriate in Attachment 2.

Sincerely, r

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Attachments c: Mr. John B. Martin Mr. William Dixon Regional Administrator, Region V State of Oregon U.S. Nuclear Regulatory Commission Department of Energy Mr. Michael J. Sykes Mr. R. C. Earr NRC Resident Inspector Chairman of County Commissioners Trojan Nuclear Plant s

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-LCA 159 Attachment-1 1Page 1 of 3 NRC Ouestions Although the proposed change is categorized as an "Administrative'" change, -

the application fails to consider:

1. . the appropriateness of Technical Specification =(TS)'3.9.4 and proposed TS-3.9.9 during those Mode 6 periods when core alterations or movement of irradiated fuel within the. containment is not in progress,'yet heavy load (eg, reactor vessel head, overhead loads) activities may be in progress over the refueling cavity; ,

2.- the absence of a TS under such conditions, which would petuit such operations to be conducted without requiring containment isolation capability;

3. the appropriateness of TS 3.9.4 and proposed.TS 3.9.9 during Mode 6 3'

operations (not core alterations),'in light of Trojan's loose fuel particle / contamination experience, and the need for containment isolation capability; and

4. the inconsistency of TS 3.9.4 with respect to TS 3.9.9 vice the reverse. ,

PGs Response

1. An analysis of the consequences of heavy load drops in the vicinity of the reactor vessel was provided in PGE to NRC letter, "Control of Heavy Loads, NUREG-0612", January 22, 1982. The heavy loads of concern were the reactor vessel head, the core upper internals, and the missile shields.

While removing the reactor vessel head, the potential exists to drop the head back onto the vessel. This drop could occur through air or a combination of air and water and could occur from various heights.

To evaluate the separate cases, an analysis was performed using con-ventional energy balance methods similar to those described in Westinghouse Topical Report WCAP-9198 "Reactor Vessel Head Drop Analyses". These Plant-specific analyses determined the consequences to the reactor vessel and its sv.pports, and verified the conclusions

reached in the Westinghouse report that "there will be no consequen-tial damage to the structural integrity of the vessel nozzles and core-coolir.g capability, and the integrity of the fuel cladding will be maintained".

A drop of the missile shields from their maximum lift height was analyzed to determine the effects on the reactor vessel and head.

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LCA 159 Attachrent 1 Page 2 of 3

-Assuming a worst-case orientation of a missile shield on impact with the head, the analysis showed that penetration of the pressure boundary would not occur. The consequences to the reactor vessel and its supports would be~1ess severe then discussed in the head drop analysis above, since the impact energies of the missile shields dropped from their maximum height are less than the energy of the worst-case ' reactor vessel head drop.

Analysis of a potential drop of the reactor vessel core upper internals into the reactor vessel was performed to determine if unacceptable damage would occur to fuel in the core or to the reactor vessel. The anL2ysis demonstrated that fuel would not be impacted and the reactor vessel would not be unacceptably damaged.

A summary'of the above' analysis is provided in Final Safety Analysis Report (FSAR) Section 9.1.5.

? Administrative Order (AO) 3-20. "Control of Heavy. Loads and Special Lif ting Devices", is the Trojan procedure used to control the lif ting of heavy loads inside Containment. The procedure precludes the lifting of heavy loads over the reactor vessel for which specific analyses were not perforzad by designating safe' load paths.

The NRC Safety Evaluation Report (SER) of July 18, 1983 on Trojan's review of heavy loads concluded that ". . . PGE has stated and defined safe load paths based on appropriate consideration for minimizing the exposure of spent fuel. . . The definition and implementation of safe load paths at the Trojan Plant provide a degree of protection from the consequences of a load drop equivalent to that contained in NUREG-0612."

The ausluation of heavy loads was performed independent of the availability of design featursa such as Containment ventilation isolation. Therefore, changes proposed to Trojan Technical Specifi-cation 3.9.9 do not alter the review, and acceptance by the NRC, of the Trojan heavy loads analysis.

2. Generic Letter 85-11. "Completion of Phase II of ' Control of Heavy Loads at Nuclear Power Plants', NUREG-0612", concluded that imple-mentation of controls and analyses for heavy loads provides sufficient protection such that risk from heavy lo&d drops is acceptably small and no further generic action is required. Based on this, and the NRC issuance of the aforementioned SER on heavy loads, the need for a Technical Specification on heavy loads is not evident.

3. The contamination that occurred at Trojan in April 1987 due to dis-crete radioactive particles was not significant enough to initiate a Containment ventilation isolation signal. Therefore, the operability

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LCA 159 Attachment 1 Page 3 of 3 of the Containment ventilation isolation system was irrelevant to this perticular event. Since that event,' extensive changes have been implemented to enhance the capability to detect and control discrete radioactive particles, which should reduce the likelihood and magni-tude of contamination from discrete radioactive particles in the future.

4. In Revisions 0 and 1 of the Westinghouse Standard Technical Specifica-tions (M-STS), NUREG-0452 Technical Specification 3.9.9 was noted as applicable in Mode 6. Trojan Technical Specifications were modelled after these versions of the M-STS. In Revision 2 of the M-STS (July

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1979), Technical Specification 3.9.9 was revised to only be applicable during core alterations or movement of irradiated fuel within he Containment. The current version of the M-STS (Revision 4) is the same as Revision 2, insofar as applicability of the Technical Specifi-cation is concerned. In addition, a review of the Technical Specifi-cations for other plants (Diablo Canyon, Wolf Creek, and Vogtle) indicates that this Technical Specification is only applicable during core alterations or movement of irradiated fuel within the Contain-ment. Therefore, it is evident that changing the applicability of Trojan Technical Specification 3.9.9 from "Mode 6" to "during core alterations or movement of irradiated fuel within the Containment",

would not only be consistent with the M-STS as issued by the NRC, but also would be consistent with the Technical Specifications issued by the NRC for other similar nuclear plants.

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LCA 159

. Att01hment 2 page 1 of 4 DgSCRIPTION OF CHANGs Trojan Technical Specification (TTS) 3.9.9 - Containment Ventilation Isolation System

- Applicability is changed from "N0DE 6" to "During CORE ALTERATIONS or movement of irradiated fuel within the Containment".

- The Action statement is changed to specify closing Containment Ventilation penetrations providing direct access from the Contain-ment atmosphere to the outside atmosphere if the Containment Ven-tilation Isolation System (CVIS) is inoperable.

REASON FOR CHANGE ,

The reasons for the change to the Applicability are to resolve an in-consistency between TTS 3.9.9 (CVIS) and TTS 3.9.4 (Containment Building penetrations) Limiting Conditions For Operation, and to resolve an incon-sistency between TTS 3.9.9 Applicability and Surveillance Requirements.

TTS 3.9.4 establishes closure requirements for Containment Building penetrations during Refueling Operations when CORE ALTERATIONS or fuel movement is in progress inside the Containment. At other times in Mode 6, the Containment Building may be opened up, with the airlocks, equipment hatch and other penetrations open.

l TTS 3.9.9 requires that each penetration providing direct access from the Containment atmosphere to the outside stmosphere be closed while in Mode 6 if the CVIS is inoperable. Since the Containment is normally open in Mode 6 as permitted by TTS 3.9.4, it is inconsistent to single out containment Ventilation dampers for closure if CVIS operability is not maintained. In addition, the Surveillance Requirements of TTS 3.9.9 are required to be perforsed only before starting and while conducting CORE ALTERATIONS. No Surveillance Requirements for Mode 6 are specified if fuel movement or CORE ALTERATIONS are not in progress. The change, thus, removes this inconsistency between TTS 3.9.9 Applicability and Surveil-lance Requirements.

j The reason for the change to the TTS 3.9.9 Action statement is to make it clear that only Containment Ventilation penetrations must be closed as a result of CVIS inoperability.

SICMIFICANT HAZARDS CONSIDERATION DETERMINATION This proposed change will revise the Applicability statement for the CVIS Refueling Operations Technical Specification. The CVIS is presently required to be operable at all times in Mode 6. This change will revise the CVIS Refueling Operations Technical Specification such that it is applicable during CORE ALTERATIONS or movement of irradiated fuel within

LCA 159

. Attcchment 2 Page 2 of 4 the Containment. Additionally, the Action statement is revised to specify closing Containment Ventilation penetrations if CVIS is inoperable.

This change does not involve a significant increase in the probability or consequences of an accident. Status of the CVIS does not affect accident probability, because this system functions to isolate Containment Venti-lation penetrations after an accident has occurred. Operability of the CVIS enables automatic isolation of the Containment Ventilation penetra-tions. Containment Building penetration closure, while in Mode.6, is relied upon only while CORE ALTERATIONS or movement of irradiated fuel  !

within the Containment is in progress per TTS 3.9.4. This Technical Specification requires the containment equipment door be closed, one door in each air lock be closed, and each penetration from Containment atmos-phere to outside atmosphere be either closed or capable of being closed by a Containment Ventilation isolation valve. These operating and clo-sure restrictions on Containment Building penetrations are sufficient to restrict radioactive material release from a fuel element rupture while in the Refueling mode. As a result of this change, the same Applica- ,

bility requirements will apply to CVIS as apply to Containment Building penetrations. The CVIS will be required to be operable to support Con-tainment Building penetration closure requirements. The change to the Action statement is an administrative change to provide clarification that only valves which rely on CVIS for automatic closure are required to be closed should CVIS become inoperable. The consequences of an accident are, thus, not increased, since this change does not involve a substan-tive change to the condition of the Containment Building in Mode 6.

This change does not create the possibility of a new or different kind of accident from any previously evaluated. The operability of CVIS is not relevant to accident creation, since CVIS functions to mitigate the con-sequences of an accident after it has occurred.

This change does not involve a significant reduction in a margin of safety. The requirements for CVIS operability will match those for Con-tainment Building penetration closure during Refueling Operations, and affected Containment Ventilation valves will be closed should CVIS become inoperable. Penetration closure and CVIS operability will be required during CORE ALTERATIONS and fuel movement in Containment. Relaxation of CVIS operability requirements will be permitted only when these evolu-tions are not in progress. The required status of the Containment Build-ing will not be materially affected during Mode 6.

An analysis of the consequences of heavy load drops in the vicinity of the reactor vessel was provided in PGE to NRC letter, "Control of Heavy Loads, NUREG-0612", January 22, 1982. The heavy loads of concern were the reactor vessel head, the core upper internals, and the missile shields. The analysis demonstrated that a drop of these loads would not result in unacceptable fuel damage. A summary of this analysis is

O LCA 159

• Attcchment 2 Page 3 of 4 provided in Final Safety Analysis Report (FSAR) Section 9.1.5.

Administrative Order (AO) 3-20, "Control of Heavy Loads and Special Lifting Devices", is the Trojan procedure used to control the lifting of heavy loads inside Containment. The procedure precludes the lif ting of heavy loads over the reactor vessel for which specific analyses were not performed by designating safe load paths. The NRC Safety Evaluation Report (SER) of July 18, 1983 on Trojan's review of heavy loads concluded that ". . . PGE has stated and defined safe load paths based on appropri-ato consideration for minimizing the exposure of spent fuel. . . The definition and implementation of safe load paths at the Trojan Plant provide a degree of protection from the consequences of a load drop equivalent to that contained in NUREG-0612."

In addition, this change will not result in a significant reduction in a margin of safety in regard to recent operating experience associated with discrete radioactive particles. The contamination that occurred at Trojan in April 1987 due to discrete radioactive particles was not sig-nificant enough to initiate a Containment ventilation isolation signal.

Therefore, the operability of the Containment ventilation isolation sys-tem was irrelevant to this particular event. Since that event, extensive changes have been implemented to enhance the capability to detect and control discrete radioactive particles, which should reduce the likeli-hood and magnitude of contamination from discrete radioactive particles in the future.

In the March 6, 1986 Federal Remister, the NRC published a list of examples of amendments that are not likely to involve significant hazards concerns. The following example from that list applies to this proposed change, demonstrating why this amendment is not considered likely to involve a significant hazards consideration:

"A purely administrative change to Technical Specifi-cations, eg, a change to achieve consistency through-out the Technical Specifications, correction of an error, or a change in nomenclature."

This change is administrative in nature to achieve consistency throughout the Technical Specifications. Additionally, the change is consistent with the regulatory guidance of NUREC-0452, Revision 4. Standard Techni-cal Specifications for Westinghouse Pressurized Water Reactors (W-STS).

TTS 3.9.4 and 3.9.9 presently have different Applicability and Action statements for very similar and interrelated Limiting Conditions For Operation (LCO). This change will make these two specifications consis-tent with each other. Additionally, the Surveillance Requirements for LCO 3.9.9 are keyed to the commencement of CORE ALTFRATIONS. Therefore, this change would make Technical Specification 3.9.9 Applicability and Action consistent with its Surveillance Requirements.

LCA 159

. Attcchment 2

.Page 4 of 4 W-STS 3.9.9 for Containment purge and exhaust isolation system (equiva-lent to TTS 3.9.9) is applicable during CORE ALTERATIONS or movement of '

irradiated fuel within the Containment. Its Action statement requires purge and exhaust penetrations providing direct access from the Contain-ment atmosphere to outside atmosphere be clossd when the isolation system is inoperable. The requirements of this proposed license change are thus equivalent to that of regulatory guidance per M-STS.

Based on the above evaluation, this change does not pose a significant hazard.

SAFETY / ENVIRONMENTAL EVALUATION Safety and environmental evaluations were performed as required by Title 10 Code of Federal Regulations, part 50 and the TTS. This review determined that the proposed change does not create an unreviewed safety question since it is administrative in nature to achieve consistency throughout the TTS.

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