ML20207L679

Forwards Emergency Operating Procedure Generation Package Upgrade for Integrated Plan Issue BN-032,per Suppl 1 to NUREG-0737.Basis for Technical Guidelines & Plant Description to Facilitate Review Also Encl

ML20207L679
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/30/1986
From:	Frisch R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20207L682	List: ML20207L679 ML20207L687 ML20207L692 ML20207L703 ML20207L717 ML20207L744
References
RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8701120306
Download: ML20207L679 (4)

Text

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consumers

@mswsmmus AIKAMEAM5 PREsAE55 Power Generet offices: 1946 West Pernell Road, Jackson, MI 49201 + (517) 788-0550 l

December 30, 1986 Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -

INTEGRATED PLAN ISSUE BN-032 - EMERGENCY OPERATING PROCEDURES UPGRADE PROCEDURE GENERATION PACKAGE As a result of Supplement 1 to NUREG 0737 (TMI Action Plan) all facilities were required to upgrade their emergency response capabilities. Two items in this upgrade are the Control Room Design Review and Emergency Operating Procedure Upgrade. Since Big Rock Point is older and considerably different in design than modern nuclear power plants this upgrade provided a unique challenge for Consumers Power Company. Because the design of the Big Rock Point Plant differs in many ways from the BWR Owners Group base plant (Hatch)

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which was used to develop the symptom oriented emergency operating procedures, it was felt that verbatim use of the generic guidelines at Big Rock Point would be inappropriate.

The increased number of. proposed modifications following the TMI accident and Consumers Power Company's interest in determining the risk to public health and safety resulting from operation of Big Rock Point led the Company to perform a plant specific Probabilistic Risk Assessment (PRA) which was reviewed and accepted by the NRC in their Safety Evaluation Report dated

-May 17, 1984. From the PRA a list of dominant accident sequences was developed. These accident sequences were used to perform the operator function and task analysis portion of the Control Room Design Review. These analyses looked at assumed equipment failures in the dominant accident sequences and developed a corresponding list of required operator actions which would mitigate the assumed failure. Thus, a table of operator actions (

versus equipment failure was created. '

While the Control Room Design Review operator function and task analysis was ongoing, a preliminary Plant-Specific Technical Guideline was developed using 8701120306 861230 PDR ADOCK 05000155 F PDR

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Director, Nuclear Reactor Regulation' 2 Big Rock Point Plant Integrated Plan Issue BN-032 December 30, 1986 the BWR Owners Group Generic Technical Guidelines as a basis. The generic guidelines were modified to accommodate the design differences between the generic plant (Hatch) and Big Rock Point. All of the areas of concern in the generic guidelines were included or dispositioned in the Big Rock Point preliminary Plant-Specific Technical Guidelines (ie, the generic guidelines have a concern with suppression pool temperature and although Big Rock Point has no suppression pool. the temperature of the containment sump is a concern for recycle purposes and was therefore addressed).

The preliminary Plant-Specific Technical Guidelines were then compared to the table developed from the Big Rock Point PRA containing the operator actions versus equipment failure as well as existing alarm response and off-normal procedures._ The guidelines were then modified to incorporate those actions which needed to be addressed by the Emergency Operating Procedures (EOPs).

The final Big Rock Point Plant-Specific Technical Guidelines contain all of the actions identified in the dominant accident sequences of the Big Rock Point PRA as well as meeting the intent of the generic guidelines which go beyond a plant's design basis. The above actions are considered to be the equivalent to a line by line description and.therefore a comparison to the generic guidelines of the Big Rock Point Plant-Specific Technical Guidelines is not necessary.

Inasmuch as many of the conditions addressed in the Plant-Specific Technical Guidelines assume more accident scenarios then those included in the plant design basis, some of the actions proposed to mitigate the consequences of these additional scenarios are not required by or included in the Big Rock Point Technical Specifications. The following is a listing of those actions which are not permitted by the plant licensing basis during normal plant operation but which are considered prudent during emergencies and are specified in the Plant-Specific Technical Guidelines:

• In the PCS Level Control guideline, opening MO-7073 and MO-7074 to add fire water to the condenser hotwell which has the potential of diverting core spray flow. (License Amendment 15 granted interim approval to use this makeup supply pending installation of a modified ring spray system.)

• In the Containment Pressure Control guideline, allowing containment pressure to increase above the design pressure of 27 psig. },

• In the Containment Level Control guideline, allowing containment water

• 1evel to increase above the design maximum elevation (596'). -

• In the Containment Level Control guideline, specifying "high pressure recycle" as a means of recycling and cooling water in the containment sump when normal use of the post incident system is not available.

This method will involve pumping water from the containment sump into the condenser hotwell via an isolated section of the fire system; OC1286-0181-NLO4 i

Director, Nuclear Reactor Regulation 3 Big Rock Point Plant Integrated Plan Issue BN-032 December 30, 1986 water would then be returned to the PCS via the feedwater/ condensate system. . Criteria for using this procedure will include the radioactivity of the water to be recycled.

• In the Containment Level Control guideline, throttling core spray flow.

• In the ATWS contingency and the PCS Pressure Control guideline, bypassing MSIV isolation interlocks to permit opening the MSIV when containment pressure is greater than 1.0 psig.

• In the ATWS contingency, inhibiting automatic operation of the Reactor Depressurization System.

• In the PCS Pressure Control guideline, use of the Reactor Depressurization System to control PCS pressure (by establishing a blowdown path through one RDS depressurization valve and the isolation valve 1-li" bypass line).

The first five attachments are the elements of the Procedure Generation Package (PGP) for the Big Rock Point upgraded Emergency Operating Procedures (EOPs) which satisfy the requirements of NUREG 0737 - Supplement 1. The last two items are included to facilitate the review process. Included are:

• Attachment 1 - E0P Training Program

• Attachment 2 - E0P Verification Procedure

• Attachment 3 - E0P Validation Procedure

• Attachment 4 - Plant-Specific Writer's Guide for E0Ps

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• Attachment 5 - Plant-Specific Technical Cuidelines

• Attachment 6 - Basis Document for Technical Guidelines

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• Attachment 7 - Big Rock Point Plant Description This submittal completes the Integrated Plan, Issue BN-032, scheduler commitment to have the PGP submitted by December 31, 1986. It is anticipated that the remaining tasks associated with upgrading the E0Ps will be completed by the end of the 1988 refueling outage, which is currently scheduled for early 1988.

l With the exception of reviewing the guidelines against the Human Engineering Deficiencies (HEDs) identified during the Control Room Design Review (CRDR),

the guidelines are considered to be complete. Review of the guidelines against the HEDs is expected to be complete by the end of August, 1987.

i Preliminary discussions with CRDR team personnel indicate that no substantive l changes to the E0Ps (or the CRDR) should result from this review. However, i some changes to the Plant-Specific Technical Guidelines may occur as a result l of the review by the Big Rock Point Plant Review Committee. This review is l scheduled to begin in January, 1987.

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OC1286-0181-NLO4 i

p m Director, Nuclear Reactor Regulation 4 Big Rock Point Plant

. Integrated Plan Issue BN-032 December 30, 1986 After the NRC has assigned a reviewer Consumers Power Company suggests that a meeting be held at the staff's convenience, preferably at the plant site, to discuss the Big Rock Point Plant-Specific Technical Guidelines. In this forum we would be able to provide additional detail on the methodology used to develop the guidelines when considering the uniqueness of the Big Rock Point ,

facility design.

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-MN hR r.lsch Senior Licensing Analyst CC Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point Plant Attachments i

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OC1286-0181-NLO4 i

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